Design and development of innovative passive valves for Nuclear Power Plant applications

Energy Technology Data Exchange (ETDEWEB)

Sapra, M.K., E-mail: sapramk@barc.gov.in; Kundu, S.; Pal, A.K.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

2015-05-15

Highlights: • Passive valves are self-acting valves requiring no external energy to function. • These valves have been developed for Advanced Heavy Water Reactor (AHWR) of India. • Passive valves are core components of passive safety systems of the reactor. • Accumulator Isolation Passive Valve (AIPV) has been developed and tested for ECSS. • AIPV provided passive isolation and flow regulation in ECCS of Integral Test Loop. - Abstract: The recent Fukushima accident has resulted in an increased need for passive safety systems in upcoming advanced reactors. In order to enhance the global contribution and acceptability of nuclear energy, proven evidence is required to show that it is not only green but also safe, in case of extreme natural events. To achieve and establish this fact, we need to design, demonstrate and incorporate reliable ‘passive safety systems’ in our advanced reactor designs. In Nuclear Power Plants (NPPs), the use of passive safety systems such as accumulators, condensing and evaporative heat exchangers and gravity driven cooling systems provide enhanced safety and reliability. In addition, they eliminate the huge costs associated with the installation, maintenance and operation of active safety systems that require multiple pumps with independent and redundant electric power supplies. As a result, passive safety systems are preferred for numerous advanced reactor concepts. In current NPPs, passive safety systems which are not participating in day to day operation, are kept isolated, and require a signal and external energy source to open the valve. It is proposed to replace these valves by passive components and devices such as self-acting valves, rupture disks, etc. Some of these innovative passive valves, which do not require external power, have been recently designed, developed and tested at rated conditions. These valves are proposed to be used for various passive safety systems of an upcoming Nuclear Power Plant being designed

Westinghouse AP1000 advanced passive plant: design features and benefits

International Nuclear Information System (INIS)

Walls, S.J.; Cummins, W.E.

2003-01-01

The Westinghouse AP1000 Program is aimed at implementing the AP1000 plant to provide a further major improvement in plant economics while maintaining the passive safety advantages established by the AP600. An objective is to retain to the maximum extent possible the plant design of the AP600 so as to retain the licensing basis, cost estimate, construction schedule, modularization scheme, and the detailed design from the AP600 program. Westinghouse and the US Nuclear Regulatory Commission staff have embarked on a program to complete Design Certification for the AP1000 by 2004. A pre-certification review phase was completed in March 2002 and was successful in establishing the applicability of the AP600 test program and AP600 safety analysis codes to the AP1000 Design Certification. On March 28, 2002, Westinghouse submitted to US NRC the AP1000 Design Control Document and Probabilistic Risk Assessment, thereby initiating the formal design certification review process. The results presented in these documents verify the safety performance of the API 000 and conformance with US NRC licensing requirements. Plans are being developed for implementation of a series of AP1000 plants in the US. Key factors in this planning are the economics of AP1000, and the associated business model for licensing, constructing and operating these new plants. Similarly plans are being developed to get the AP1000 design reviewed for use in the UK. Part of this planning has been to examine the AP1000 design relative to anticipated UK safety and licensing issues. (author)

EP1000 passive plant description

International Nuclear Information System (INIS)

Saiu, G.

1999-01-01

In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In Phase I of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) will be completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. The second part, 'Phase 2B', includes both the analyses and evaluations required to demonstrate the adequacy of the design, and to support the preparation of Safety Case Report. The second part of Phase 2 of the program will start at the beginning of 1999 and will be completed in the 2001. Incorporation of the EUR has been a key design requirement for the EP1000 from the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. This paper integrates and updates the plant description reported in the IAEA TECDOC-968. The most significant developments of the EP1000 plant design during Phase 2A of the EPP program are described and reference is made to the key design requirements set by the EUR Rev. B document. (author)

Passive autocatalytic recombiners for combustible gas control in advanced light water reactors

International Nuclear Information System (INIS)

Wolff, U.; Sliter, G.

2004-01-01

A key aspect of the worldwide effort to develop advanced nuclear power plants is designing to address severe accident phenomena, including the generation of hydrogen during core melt progression (metal-water and core-concrete reactions). This design work not only resolves safety concerns with hydrogen, but also supports the development of a technical basis for simplification of off-site emergency planning. The dominant challenge to any emergency planning approach is a large, early containment failure due to pressure excursions. Among the potential contributors to large and rapid increases in containment pressure is hydrogen combustion. The more improbable a containment-threatening combustion becomes, the more appropriate the argument for significant emergency planning simplification. As discussed in this paper, catalytic recombiners provide a means to passively and reliably limit hydrogen combustion to a continuous oxidation process with virtually no potential for containment failure in passive advanced light water reactors (ALWRs). (author)

Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

International Nuclear Information System (INIS)

Huang Xueqing; Zhang Senru

1999-01-01

Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

Advances in passive-remote and extractive Fourier transform infrared spectroscopic systems

International Nuclear Information System (INIS)

Demirgian, J.C.; Hammer, C.; Hwang, E.; Mao, Zhuoxiong.

1993-01-01

The Clean Air Act of 1990 requires the monitoring of air toxics including those from incinerator emissions. Continuous emission monitors (CEM) would demonstrate the safety of incinerators and address public concern about emissions of hazardous organic compounds. Fourier transform infrared (FTIR) spectroscopy can provide the technology for continuous emission monitoring of stacks. Stack effluent can be extracted and analyzed in less than one minute with conventional FTIR spectrometers. Passive-remote FTIR spectrometers can detect certain emission gases over 1 km away from a stack. The authors discuss advances in both extractive and passive-remote FTIR technology. Extractive systems are being tested with EPA protocols, which will soon replace periodic testing methods. Standard operating procedures for extractive systems are being developed and tested. Passive-remote FTIR spectrometers have the advantage of not requiring an extracted sample; however, they have less sensitivity. We have evaluated the ability of commercially available systems to detect fugitive plumes and to monitor carbon monoxide at a coal-fired power plant

Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

Energy Technology Data Exchange (ETDEWEB)

Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

2013-10-15

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

Directory of Open Access Journals (Sweden)

VIKAS JAIN

2013-10-01

Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

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

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)

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

Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

International Nuclear Information System (INIS)

Karameldin, A.; Samer S. Mekhemar

2004-01-01

Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

Passive safety features for next generation CANDU power plants

International Nuclear Information System (INIS)

Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

1989-01-01

CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

EP 1000 -The European Passive Plant

International Nuclear Information System (INIS)

Cummins, Ed; Oyarzabal, Mariano; Saiu, Gianfranco

1998-01-01

A group of European utilities, along with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT) initiated a program to evaluate Westinghouse passive nuclear plant technology for application in Europe. The European utility group consisted of: Agrupacion electrica para al Desarrollo Technologico Nuclear (DTN), Spain; Electricite de France; ENEL, SpA., Italy; IVO Power Engineering, Ltd., Finland; Scottish Nuclear Limited (acting for itself on behalf of Nuclear Electric plc, U.K.; Tractebel Energy Engineering, Belgium; UAK (represented by NOK-Beznau), Switzerland; and Vattenfall AB, Ringhals, Sweden. The European Passive Plant (EPP) program, which began in 1994, is an evaluation of the Westinghouse 600 MWe AP 600 and 1000 MWe Simplified Pressurized Water Reactor (SPWR) designs in meeting the European Utility Requirements (EUR), and where necessary, modifying the design to achieve compliance. Phase 1 or the EPP program was completed and included the two major tasks of evaluating the effect of the EUR on the Westinghouse nuclear island and developing the EP 1000, a 1000 MWe passive plant reference design that conforms to the EUR and would be licensable in Europe. The EP 1000 closely follows the Westinghouse SPWR design for safety systems and containment and the AP 600 design for auxiliary systems. It also includes features that where required to meet the EUR and key European licensing requirements. The primary circuit of the EP 1000 retains most of the general features of the current-day designs, but some evolutionary features to enhance reliability, simplicity of operation, ease of maintenance, and plant safety have been incorporated into the design. The core, reactor vessel, and reactor internals of the EP 1000 are similar to those of currently operating Westinghouse PWR plants, but several new features are included to enhance the performance characteristics. The basic EP 1000 safety philosophy is based on use of inherent

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

Balancing passive and active systems for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

1999-01-01

Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

High-inertia hermetically sealed main coolant pump for next generation passive nuclear power plants

International Nuclear Information System (INIS)

Kujawski, Joseph M.; Nair, Bala R.; Vijuk, Ronald P.

2003-01-01

The main coolant pump for the Westinghouse AP1000 advanced passive nuclear power plant represents a significant scale-up in power, flow capacity, and physical size from its predecessor designed for the smaller AP600 power plant. More importantly, the AP1000 pump incorporates several innovative features that contribute to improved efficiency, operational reliability, and plant safety. The features include an internals design which provides the highest hydraulic efficiency achieved in commercial nuclear power plant applications. Another feature is the use of a distributed inertial mass system in the rotating assembly to develop the high rotational inertia to meet the extended system flow coastdown requirement for core heat removal in the event of loss of power to the pumps. This advanced canned motor pump also incorporates the latest development in higher operating voltage, providing plant designers with the ability to eliminate plant transformers and operate directly on the site electrical bus in many cases. The salient features of the pump design and performance data are presented in this paper. (author)

Design criteria for the electrical system in advanced passive reactors. Special features of the AP-600 Reactor

International Nuclear Information System (INIS)

Moraleda Lopez, A.

1997-01-01

The design of the electrical system of an Passive Advanced Reactor is determined by the concept of passive actuation of safety systems, simplification of process systems and optimisation of equipment performance. The system that results from these criteria is very different to those designed for present plants. The main differences are: No class 1E alternating current systems No emergency diesel generators Fewer safety and non-safety class electricity consumers System for continuous monitoring of battery status Use of electronic speed regulators for reactor feedwater pump motors Outsite battery backup safety power supply Motor-operated valves are the only safety electrical actuators Portable power supply for post 72 hour equipment This paper develops these concepts and applies them to the AP-600 project and describes the electrical system of this type of plant. (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.)

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Plant maintenance and advanced reactors issue, 2008

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal [ed.

2009-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

Plants for passive cooling. A preliminary investigation of the use of plants for passive cooling in temperate humid climates

Energy Technology Data Exchange (ETDEWEB)

Spirn, A W; Santos, A N; Johnson, D A; Harder, L B; Rios, M W

1981-04-01

The potential of vegetation for cooling small, detached residential and commercial structures in temperate, humid climates is discussed. The results of the research are documented, a critical review of the literature is given, and a brief review of energy transfer processes is presented. A checklist of design objectives for passive cooling, a demonstration of design applications, and a palette of selected plant species suitable for passive cooling are included.

Passive Flora? Reconsidering Nature’s Agency through Human-Plant Studies (HPS

Directory of Open Access Journals (Sweden)

John Charles Ryan

2012-08-01

Full Text Available Plants have been—and, for reasons of human sustenance and creative inspiration, will continue to be—centrally important to societies globally. Yet, plants—including herbs, shrubs, and trees—are commonly characterized in Western thought as passive, sessile, and silent automatons lacking a brain, as accessories or backdrops to human affairs. Paradoxically, the qualities considered absent in plants are those employed by biologists to argue for intelligence in animals. Yet an emerging body of research in the sciences and humanities challenges animal-centred biases in determining consciousness, intelligence, volition, and complex communication capacities amongst living beings. In light of recent theoretical developments in our understandings of plants, this article proposes an interdisciplinary framework for researching flora: human-plant studies (HPS. Building upon the conceptual formations of the humanities, social sciences, and plant sciences as advanced by Val Plumwood, Deborah Bird Rose, Libby Robin, and most importantly Matthew Hall and Anthony Trewavas, as well as precedents in the emerging areas of human-animal studies (HAS, I will sketch the conceptual basis for the further consideration and exploration of this interdisciplinary framework.

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

Many of the advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive rather than active systems to perform safety functions. Despite the reduced redundancy of the passive systems as compared to active systems in current plants, the assertion is that the overall safety of the plant is enhanced due to the much higher expected reliability of the passive systems. In order to investigate this assertion, a study is being conducted at Sandia National Laboratories to evaluate the reliability of ALWR passive safety features in the context of probabilistic risk assessment (PRA). The purpose of this paper is to provide a brief overview of the approach to this study. The quantification of passive system reliability is not as straightforward as for active systems, due to the lack of operating experience, and to the greater uncertainty in the governing physical phenomena. Thus, the adequacy of current methods for evaluating system reliability must be assessed, and alternatives proposed if necessary. For this study, the Westinghouse Advanced Passive 600 MWe reactor (AP600) was chosen as the advanced reactor for analysis, because of the availability of AP600 design information. This study compares the reliability of AP600 emergency cooling system with that of corresponding systems in a current generation reactor

Utility requirements for safety in the passive advanced light-water reactor

International Nuclear Information System (INIS)

Marston, T.U.; Layman, W.H.; Bockhold, G. Jr.

1993-01-01

The objective of the passive plant design is to use passive systems to replace all the active engineered safety systems presently used in light-water reactors. The benefits derived from such an approach to safety design are multiple. First, it is expected that a passive design approach will significantly simplify the overall plant design, including a reduction in the number of components, and reduce the operation and maintenance burden. Second, it is expected that the overall safety and reliability of the passive systems will be improved over active systems, which will result in extremely low risk to public health and safety. Third, challenges to the operating staff will be minimized during transient and emergency conditions, which will reduce the uncertainty associated with human behavior. Finally, it is expected that reliance on passive safety features will lead to a better understanding by the general public and recognition that a major improvement in public safety has been achieved

NRC review of Electric Power Research Institute's advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

International Nuclear Information System (INIS)

1994-08-01

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the open-quotes Advanced Light Water Reactor [ALWR] Utility Requirements Documentclose quotes, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, open-quotes ALWR Policy and Summary of Top-Tier Requirementsclose quotes, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, open-quotes NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summaryclose quotes, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review

Advanced configuration of hybrid passive filter for reactive power and harmonic compensation.

Science.gov (United States)

Kececioglu, O Fatih; Acikgoz, Hakan; Sekkeli, Mustafa

2016-01-01

Harmonics is one of the major power quality problems for power systems. The harmonics can be eliminated by power filters such as passive, active, and hybrid. In this study, a new passive filter configuration has been improved in addition to the existing passive filter configurations. Conventional hybrid passive filters are not successful to compensate rapidly changing reactive power demand. The proposed configure are capable of compensating both harmonics and reactive power at the same time. Simulation results show that performance of reactive power and harmonic compensation with advanced hybrid passive filter is better than conventional hybrid passive filters.

European passive plant program A design for the 21st century

International Nuclear Information System (INIS)

Adomaitis, D.; Oyarzabal, M.

1998-01-01

In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The following major tasks were accomplished: (1) the impacts of the European utility requirements (EUR) on the Westinghouse nuclear island design were evaluated; and (2) a 1000 MWe passive plant reference design (EP1000) was established which conforms to the EUR and is expected to be licensable in Europe. With respect to safety systems and containment, the reference plant design closely follows that of the Westinghouse simplified pressurized water reactor (SPWR) design, while the AP600 plant design has been taken as the basis for the EP1000 reference design in the auxiliary system design areas. However, the EP1000 design also includes features required to meet the EUR, as well as key European licensing requirements. (orig.)

ESBWR passive heat exchanger design and performance - reducing plant development costs

International Nuclear Information System (INIS)

Lumini, E.; Upton, H.A.; Billig, P.F.; Masoni, P.

1996-01-01

The EUROPEAN Simplified Boiling Water Reactor (ESBWR) is a nuclear plant that builds on the solid technological foundation of the Simplified Boiling Reactor (SBWR) design. The major objective of the ESBWR program is to develop a plant design that utilizes the basic simplicity of the SBWR design that utilizes the basic simplicity of the SBWR design features to improve overall economics and to meet the specific requirements found in the European Utility Requirements Documents (EUR). The design is being developed by an international team of utilities, designers and researchers with the objective of meeting European utility and regulatory requirements. The overall approach to improve the commercial attractiveness of the ESBWR compared to the SBWR was to take advantage of the modular design of the passive safety system, the economy of scale, as well as the advantage of simpler systems of the passive plant to reduce overall material quantities and improve plant economics. To take advantage of the economy of scale, the power level of ESBWR was increased to 1190 MWe. Because of the modular nature of the passive safety systems in SBWR, in increase in thermal power of ESBWR to 3613 MWt only requires that the number of Passive Containment Condensers to maintain the passive safety features of ESBWR to four 33 MWt units for ESBWR. This paper reviews the Passive Containment Cooling (PCC) and Isolation Condenser (IC) unit design and addresses their use in the passive safety systems of the 3613 MWt ESBWR. The specific design differences and the applicability of the test completed at the SIET PANTHERS test facility in Piacenza, Italy are addressed as well as outlining additional qualification tests that must be completed on the PCC and IC unit design if they are to used in the passive safety systems of the ESBWR. This paper outlines the test results obtained from the prototype PCC and IC PANTHERS tests facility in Piacenza, Italy which have been used to design the ESBWR PCC/1C

IAEA activities in technology development for advanced water-cooled nuclear power plants

International Nuclear Information System (INIS)

Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

2003-01-01

As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

Directory of Open Access Journals (Sweden)

Matthew Bucknor

2017-03-01

Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

Advanced reactor passive system reliability demonstration analysis for an external event

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

2017-03-15

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

Advanced reactor passive system reliability demonstration analysis for an external event

International Nuclear Information System (INIS)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

2017-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event

Reliability assurance for regulation of advanced reactors

International Nuclear Information System (INIS)

Fullwood, R.; Lofaro, R.; Samanta, P.

1992-01-01

The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. this paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

Reliability assurance for regulation of advanced reactors

International Nuclear Information System (INIS)

Fullwood, R.; Lofaro, R.; Samanta, P.

1991-01-01

The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics

Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

International Nuclear Information System (INIS)

Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young

2013-01-01

Highlights: • We newly propose the design concept of integrated passive safety system (IPSS). • It has five safety functions for decay heat removal and severe accident mitigation. • Simulations for IPSS show that core melt does not occur in accidents with SBO. • IPSS can achieve the passive in-vessel retention and ex-vessel cooling strategy. • The applicability of IPSS is high due to the installation outside the containment. -- Abstract: The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the

Liquid and solid rad waste treatment in advanced nuclear power plants. Application to the SBWR design

International Nuclear Information System (INIS)

Tielas Reina, M.; Asuar Alonso, O.

1994-01-01

Rad waste treatment requirements for the new generation of American advanced passive and evolutionary power plants are listed in the URD (Utility Requirements Document) of the EPRI (Electrical Power Research Institute). These requirements focus on: - Minimization of shipped solid wastes - Minimization of liquid effluents - Simplification of design and operation, with emphasis not only on waste treatment system design but also on general plant design and operation These objectives are aimed at: - Reducing and segregating wastes at source - Minimizing chemical contamination of these wastes System design simplification is completed by providing free space in the building for the use of mobile plants, either for special services not considered in the basic design or to accommodate future technical advances. (Author)

77 FR 56241 - Notice of Withdrawal of Final Design Approval; Westinghouse Electric Company; Advanced Passive 1000

Science.gov (United States)

2012-09-12

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0131] Notice of Withdrawal of Final Design Approval; Westinghouse Electric Company; Advanced Passive 1000 By letter dated December 10, 2010, Westinghouse Electric... final design approval (FDA) for the Advanced Passive 1000 (AP1000) design upon the completion of...

Information on the Advanced Plant Experiment (APEX) Test Facility

International Nuclear Information System (INIS)

Smith, Curtis Lee

2015-01-01

The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, 'Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing,' Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

From dodewaard to a modern economic passive plant-ESBWR

International Nuclear Information System (INIS)

Gonzalez Lopez, A.; Arnold, H.; Yadigaroglu, G.; Stoop, P.M.; Rao, A.

1997-01-01

For over 25 years the Dodewaard nuclear plant has produced electricity with one of the highest reliability of any power plant in the world. Almost 10 years ago when some designers looked at the features to incorporate in a modern mid-size BWR design, they chose some of the key features of Dodewaard natural circulation and passive safety. Over the years this design evolved into a 670 SBWR design that was developed by an international team and consisted of natural circulation and passive safety injection and decay heat removal. Since the passive decay heat removal was a major new technology area, an extensive test program was developed and conducted utilizing newly constructed large scale integrated system test facility at the Paul Scherrer Institute, Switzerland. The development of any modern reactor is time consuming and expensive, hence, the design and technology was done as part of an international team effort. This paper provides an overview of the international design and technology effort and discusses how the development costs were minimized through cooperation. (Author)

Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

International Nuclear Information System (INIS)

Winters, J.W.; Corletti, M.M.; Hayashi, Y.

2003-01-01

A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States

Role of passive valves & devices in poison injection system of advanced heavy water reactor

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

2014-01-01

The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water (H 2 O) cooled, heavy water (D 2 O) moderated reactor. The reactor design is based on well-proven water reactor technologies and incorporates a number of passive safety features such as natural circulation core cooling; direct in-bundle injection of light water coolant during a Loss of Coolant Accident (LOCA) from Advanced Accumulators and Gravity Driven Water Pool by passive means; Passive Decay Heat Removal using Isolation Condensers, Passive Containment Cooling System and Passive Containment Isolation System. In addition to above, there is another passive safety system named as Passive Poison Injection System (PPIS) which is capable of shutting down the reactor for a prolonged time. It is an additional safety system in AHWR to fulfill the shutdown function in the event of failure of wired shutdown systems i.e. primary and secondary shut down systems of the reactor. When demanded, PPIS injects the liquid poison into the moderator by passive means using passive valves and devices. On increase of main heat transport (MHT) system pressure beyond a predetermined value, a set of rupture disks burst, which in-turn actuate the passive valve. The opening of passive valve initiates inrush of high pressure helium gas into poison tanks to push the poison into the moderator system, thereby shutting down the reactor. This paper primarily deals with design and development of Passive Poison Injection System (PPIS) and its passive valves & devices. Recently, a prototype DN 65 size Poison Injection Passive Valve (PIPV) has been developed for AHWR usage and tested rigorously under simulated conditions. The paper will highlight the role of passive valves & devices in PPIS of AHWR. The design concept and test results of passive valves along with rupture disk performance will also be covered. (author)

Results of a Demonstration Assessment of Passive System Reliability Utilizing the Reliability Method for Passive Systems (RMPS)

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia; Grelle, Austin

2015-04-26

Advanced small modular reactor designs include many advantageous design features such as passively driven safety systems that are arguably more reliable and cost effective relative to conventional active systems. Despite their attractiveness, a reliability assessment of passive systems can be difficult using conventional reliability methods due to the nature of passive systems. Simple deviations in boundary conditions can induce functional failures in a passive system, and intermediate or unexpected operating modes can also occur. As part of an ongoing project, Argonne National Laboratory is investigating various methodologies to address passive system reliability. The Reliability Method for Passive Systems (RMPS), a systematic approach for examining reliability, is one technique chosen for this analysis. This methodology is combined with the Risk-Informed Safety Margin Characterization (RISMC) approach to assess the reliability of a passive system and the impact of its associated uncertainties. For this demonstration problem, an integrated plant model of an advanced small modular pool-type sodium fast reactor with a passive reactor cavity cooling system is subjected to a station blackout using RELAP5-3D. This paper discusses important aspects of the reliability assessment, including deployment of the methodology, the uncertainty identification and quantification process, and identification of key risk metrics.

From Dodewaard to a modern economic passive plant-ESBWR

International Nuclear Information System (INIS)

Arnold, H.; Yadigaroglu, G.; Stoop, P.M.; Gonzales, A.; Rao, A.S.

1996-01-01

For over 25 years the Dodewaard nuclear plant has produced electricity with one of the highest reliability of any nuclear power plant in the world. Almost 10 years ago when some designers looked at the features to incorporate in a modern mid-size Boiling Water Reactor (SBWR) design that was developed by an international team and consisted of natural circulation and passive safety injection and decay heat removal. Since the passive decay heat removal was a major new technology area, an extensive test program was developed and conducted - utilizing newly constructed large scale integrated system test facility at the Paul Scherrer Institute, Switzerland. The development of any modern reactor is time consuming and expensive, hence, the design and technology was done as part of an international team effort. This paper provides an overview of the international design and technology effort and discusses how the development costs were minimized through cooperation. The result of the design effort showed that overall plant economics could be improved by a power level increase, supported by some modest design changes. This paper provides the approach and key results of this multi-year international design effort to develop a modern, larger scale, passive plant - the European Simplified Boiling Water Reactor (ESBWR). This effort, in addition to developing an economic nuclear power station, also demonstrates how the benefits of international cooperation can be utilized to keep development costs reasonable. An overview is also provided of the planned research efforts to extend existing technology to the ESBWR. (authors)

Activities of passive cooling applications and simulation of innovative nuclear power plant design

International Nuclear Information System (INIS)

Aglar, F.; Tanrykut, A.

2002-01-01

This paper gives a general insight on activities of the Turkish Atomic Energy Authority (TAEA) concerning passive cooling applications and simulation of innovative nuclear power plant design. The condensation mode of heat transfer plays an important role for the passive heat removal application in advanced water-cooled reactor systems. But it is well understood that the presence of noncondesable (NC) gases can greatly inhibit the condensation process due to the build up of NC gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of NC. The test matrix of the experimental investigation undertaken at the METU-CTF test facility (Middle East Technical University, Ankara) covers the range of parameters; Pn (system pressure) : 2-6 bar, Rev (vapor Reynolds number): 45,000-94,000, and Xi (air mass fraction): 0-52%. This experimental study is supplemented by a theoretical investigation concerning the effect of mixture flow rate on film turbulence and air mass diffusion concepts. Recently, TAEA participated to an international standard problem (OECD ISP-42) which covers a set of simulation of PANDA test facility (Paul Scherrer Institut-Switzerland) for six different phases including different natural circulation modes. The concept of condensation in the presence of air plays an important role for performance of heat exchangers, designed for passive containment cooling, which in turn affect the natural circulation behaviour in PANDA systems. (author)

Passivity Enhancement in RES Based Power Plant with Paralleled Grid-Connected Inverters

DEFF Research Database (Denmark)

Bai, Haofeng; Wang, Xiongfei; Blaabjerg, Frede

2016-01-01

Harmonic instability is threatening the operation of power plants with multiple grid connected converters in parallel. To analyze and improve the stability of the grid connected converters, the passivity of the output admittance converters is first analyzed in this paper. It is shown that the non-passivity...

High-Throughput Phenotyping of Wheat and Barley Plants Grown in Single or Few Rows in Small Plots Using Active and Passive Spectral Proximal Sensing

Directory of Open Access Journals (Sweden)

Gero Barmeier

2016-11-01

Full Text Available In the early stages of plant breeding, breeders evaluate a large number of varieties. Due to limited availability of seeds and space, plot sizes may range from one to four rows. Spectral proximal sensors can be used in place of labour-intensive methods to estimate specific plant traits. The aim of this study was to test the performance of active and passive sensing to assess single and multiple rows in a breeding nursery. A field trial with single cultivars of winter barley and winter wheat with four plot designs (single-row, wide double-row, three rows, and four rows was conducted. A GreenSeeker RT100 and a passive bi-directional spectrometer were used to assess biomass fresh and dry weight, as well as aboveground nitrogen content and uptake. Generally, spectral passive sensing and active sensing performed comparably in both crops. Spectral passive sensing was enhanced by the availability of optimized ratio vegetation indices, as well as by an optimized field of view and by reduced distance dependence. Further improvements of both sensors in detecting the performance of plants in single rows can likely be obtained by optimization of sensor positioning or orientation. The results suggest that even in early selection cycles, enhanced high-throughput phenotyping might be able to assess plant performance within plots comprising single or multiple rows. This method has significant potential for advanced breeding.

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

European passive plant program preliminary safety analyses to support system design

International Nuclear Information System (INIS)

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

1999-01-01

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

The european passive plant (EPP) design: compliance with the european utilities requirements (EUR)

International Nuclear Information System (INIS)

Noviello, L.; Oyarzabal, M.

1996-01-01

Back 1986, most of the European firms have participated to the American program called the Advanced Light Water Reactors (ALWR) including the development of the Utilities Requirements as well as four projects as for instance AP600. Later, in the year 1990, seven European firms have begun to develop the European Utilities Requirements. This development is justified by the fact that the lessons learned by the nuclear power plants designs programs of the years 1980 can be incorporated and the European specific conditions can be taken into consideration. Thus, in 1994, eight European firms - Westinghouse and their industrial partners - have decided to launch a multiphase program in order to check the AP600 compliance with the European Utilities Requirements (EUR) and to develop the required alterations. Today, the phase I of the EPP (European Passive Plant) program has been completed. In this phase, the main important objectives have been reached. (O.M.)

Passive cooling applications for nuclear power plants using pulsating steam-water heat pipes

International Nuclear Information System (INIS)

Aparna, J.; Chandraker, D.K.

2015-01-01

Gen IV reactors incorporate passive principles in their system design as an important safety philosophy. Passive safety systems use inherent physical phenomena for delivering the desired safe action without any external inputs or intrusion. The accidents in Fukushima have renewed the focus on passive self-manageable systems capable of unattended operation, for long hours even in extended station blackout (SBO) and severe accident conditions. Generally, advanced reactors use water or atmospheric air as their ultimate heat sink and employ passive principles in design for enhanced safety. This paper would be discussing the experimental results on pulsating steam water heat-pipe devices and their applications in passive cooling. (author)

Simulation of fission products behavior in severe accidents for advanced passive PWR

International Nuclear Information System (INIS)

Tong, L.L.; Huang, G.F.; Cao, X.W.

2015-01-01

Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

Regulatory Considerations for the Long Term Cooling Safe Shutdown Requirements of the Passive Residual Heat Removal Systems in Advanced Reactors

International Nuclear Information System (INIS)

Sim, S. K.; Bae, S. H.; Kim, Y. S.; Hwang, Min Jeong; Bang, Young Seok; Hwang, Taesuk

2016-01-01

USNRC approved safe shutdown at 215.6 .deg. C for a safe and long term cooling state for the redundant passive RHRSs by SECY-94-084. USNRC issued COLA(Combined Construction and Operating License) for the Levy County NP Unit-1/2 for the AP1000 passive RHRSs in 2014. Korea Hydro and Nuclear Power(KHNP) is developing APR+ and adopted Passive Auxiliary Feedwater System(PAFS) as a new passive RHRS design. Korea Institute of Nuclear Safety(KINS) has been developing regulatory guides for the advanced safety design features of the advanced ALWRs which has plan to construct in near future in Korea[5]. Safety and regulatory issues as well as the safe shut down requirements of the passive RHRS are discussed and considerations in developing regulatory guides for the passive RHRS are presented herein. Passive RHRSs have been introduced as new safety design features for the advanced reactors under development in Korea. These passive RHRSs have potential advantages over existing active RHRS, however, their functions are limited due to inherent ability of passive heat removal processes. It is high time to evaluate the performance of the passive PRHRs and develop regulatory guides for the safety as well as the performance analyses of the passive RHRS

Preliminary assessment of a combined passive safety system for typical 3-loop PWR CPR1000

Energy Technology Data Exchange (ETDEWEB)

Yang, Zijiang; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli

2017-03-15

Highlights: • A combined passive safety system was placed on a typical 3-loop PWR CPR1000. • Three accident analyses show the three different accident mitigation methods of the passive safety system. • The three mitigation methods were proved to be useful. - Abstract: As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have developed. Thoughts have been given to upgrade the existing reactors with passive systems to meet stricter safety demands. In this paper, the CPR1000 plant, which is one kind of mature pressurized water reactor plants in China, is improved with some passive systems to enhance safety. The passive systems selected are as follows: (1) the reactor makeup tank (RMT); (2) the advanced accumulator (A-ACC); (3) the in-containment refueling water storage tank (IRWST); (4) the passive emergency feed water system (PEFS), which is installed on the secondary side of SGs; (5) the passive depressurization system (PDS). Although these passive components is based on the passive technology of some advanced reactors, their structural and trip designs are adjusted specifically so that it could be able to mitigate accidents of the CPR1000. Utilizing the RELAP5/MOD3.3 code, accident analyses (small break loss of coolant accident, large break loss of coolant accident, main feed water line break accident) of this improved CPR1000 plant were presented to demonstrate three different accident mitigation methods of the safety system and to test whether the passive safety system preformed its function well. In the SBLOCA, all components of the passive safety system were put into work sequentially, which prevented the core uncover. The LBLOCA analysis illustrates the contribution of the A-ACCs whose small-flow-rate injection can control the maximum cladding

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

Advanced power plant materials, design and technology

Energy Technology Data Exchange (ETDEWEB)

Roddy, D. (ed.) [Newcastle University (United Kingdom). Sir Joseph Swan Institute

2010-07-01

The book is a comprehensive reference on the state of the art of gas-fired and coal-fired power plants, their major components and performance improvement options. Selected chapters are: Integrated gasification combined cycle (IGCC) power plant design and technology by Y. Zhu, and H. C. Frey; Improving thermal cycle efficiency in advanced power plants: water and steam chemistry and materials performance by B. Dooley; Advanced carbon dioxide (CO{sub 2}) gas separation membrane development for power plants by A. Basile, F. Gallucci, and P. Morrone; Advanced flue gas cleaning systems for sulphur oxides (SOx), nitrogen oxides (NOx) and mercury emissions control in power plants by S. Miller and B.G. Miller; Advanced flue gas dedusting systems and filters for ash and particulate emissions control in power plants by B.G. Miller; Advanced sensors for combustion monitoring in power plants: towards smart high-density sensor networks by M. Yu and A.K. Gupta; Advanced monitoring and process control technology for coal-fired power plants by Y. Yan; Low-rank coal properties, upgrading and utilisation for improving the fuel flexibility of advanced power plants by T. Dlouhy; Development and integration of underground coal gasification (UCG) for improving the environmental impact of advanced power plants by M. Green; Development and application of carbon dioxide (CO{sub 2}) storage for improving the environmental impact of advanced power plants by B. McPherson; and Advanced technologies for syngas and hydrogen (H{sub 2}) production from fossil-fuel feedstocks in power plants by P. Chiesa.

Soil Seed Bank and Plant Community Development in Passive Restoration of Degraded Sandy Grasslands

Directory of Open Access Journals (Sweden)

Renhui Miao

2016-06-01

Full Text Available To evaluate the efficacy of passive restoration on soil seed bank and vegetation recovery, we measured the species composition and density of the soil seed bank, as well as the species composition, density, coverage, and height of the extant vegetation in sites passively restored for 0, 4, 7, and 12 years (S0, S4, S7, and S12 in a degraded grassland in desert land. Compared with S0, three more species in the soil seed bank at depths of 0–30 cm and one more plant species in the community was detected in S12. Seed density within the topsoil (0–5 cm was five times higher in S12 than that in S0. Plant densities in S7 and S12 were triple and quadruple than that in S0. Plant coverage was increased by 1.5 times (S4, double (S7, and triple (S12 compared with S0. Sørensen’s index of similarity in species composition between the soil seed bank and the plant community were high (0.43–0.63, but it was lower in short-term restoration sites (S4 and S7 than that in no and long-term restoration sites (S0 and S12. The soil seed bank recovered more slowly than the plant community under passive restoration. Passive restoration is a useful method to recover the soil seed bank and vegetation in degraded grasslands.

Comparing herbaceous plant communities in active and passive riparian restoration.

Directory of Open Access Journals (Sweden)

Elise S Gornish

Full Text Available Understanding the efficacy of passive (reduction or cessation of environmental stress and active (typically involving planting or seeding restoration strategies is important for the design of successful revegetation of degraded riparian habitat, but studies explicitly comparing restoration outcomes are uncommon. We sampled the understory herbaceous plant community of 103 riparian sites varying in age since restoration (0 to 39 years and revegetation technique (active, passive, or none to compare the utility of different approaches on restoration success across sites. We found that landform type, percent shade, and summer flow helped explain differences in the understory functional community across all sites. In passively restored sites, grass and forb cover and richness were inversely related to site age, but in actively restored sites forb cover and richness were inversely related to site age. Native cover and richness were lower with passive restoration compared to active restoration. Invasive species cover and richness were not significantly different across sites. Although some of our results suggest that active restoration would best enhance native species in degraded riparian areas, this work also highlights some of the context-dependency that has been found to mediate restoration outcomes. For example, since the effects of passive restoration can be quite rapid, this approach might be more useful than active restoration in situations where rapid dominance of pioneer species is required to arrest major soil loss through erosion. As a result, we caution against labeling one restoration technique as better than another. Managers should identify ideal restoration outcomes in the context of historic and current site characteristics (as well as a range of acceptable alternative states and choose restoration approaches that best facilitate the achievement of revegetation goals.

Passive Solar still: Recent advancement in design and related Performance.

Science.gov (United States)

Awasthi, Anuradha; Kumari, Kanchan; Panchal, Hitesh; Sathyamurthy, Ravishankar

2018-05-31

Present review paper mainly focuses on different varieties of solar stills and highlights mostly the passive solar still with advanced modifications in the design and development of material, single and multi-effect solar still with augmentation of different materials, energy absorbing, insulators, mechanisms of heat and mass transfer to improve the loss of heat and enhance the productivity of solar still. The cost-benefit analysis along with the progressive advancement for solar stills is the major highlights of this review. To increase the output of solar still nowadays, applications of advance modifications is one of the promising tools, and it is anticipated that shortly more vigor will be added in this area with the modifications in designs of solar stills.

Use of passive systems to improve plant operation and maintenance

International Nuclear Information System (INIS)

Shah, D.

2000-01-01

In a deregulated future, a utility's strength will depend on its ability to be cost competitive in the marketplace. However, the competitive advantage of nuclear power will depend on each owner's ability to reduce Operating and Maintenance (O and M) costs without sacrificing nuclear safety. The use of passive systems (i.e., systems without any moving parts) can reduce plant O and M costs while increasing safety in nuclear power plants. (author)

Advances in passive neutron instruments for safeguards use

International Nuclear Information System (INIS)

Menlove, H.O.; Krick, M.S.; Langner, D.G.; Miller, M.C.; Stewart, J.E.

1994-01-01

Passive neutron and other nondestructive assay techniques have been used extensively by the International Atomic Energy Agency to verify plutonium metal, powder, mixed oxide, pellets, rods, assemblies, scrap, and liquids. Normally, the coincidence counting rate is used to measure the 240 Pu-effective mass and gamma-ray spectrometry or mass spectrometry is used to verify the plutonium isotopic ratios. During the past few years, the passive neutron detectors have been installed in plants and operated in the unattended/continuous mode. These radiation data with time continuity have made it possible to use the totals counting rate to monitor the movement of nuclear material. Monte Carlo computer codes have been used to optimize the detector designs for specific applications. The inventory sample counter (INVS-III) has been designed to have a higher efficiency (43%) and a larger uniform counting volume than the original INVS. Data analyses techniques have been developed, including the ''known alpha'' and ''known multiplication'' methods that depend on the sample. For scrap and other impure or poorly characterized samples, we have developed multiplicity counting, initially implemented in the plutonium scrap multiplicity counter. For large waste containers such as 200-L drums, we have developed the add-a-source technique to give accurate corrections for the waste-matrix materials. This paper summarizes recent developments in the design and application of passive neutron assay systems

Experimental research on passive residual heat remove system for advanced PWR

International Nuclear Information System (INIS)

Huang Yanping; Zhuo Wenbin; Yang Zumao; Xiao Zejun; Chen Bingde

2003-01-01

The experimental and qualified results of MISAP in the research of passive residual heat remove system of advanced PWR performed in the Bubble physics and natural circulation laboratory in Nuclear Power Institute of China in the past ten years is overviewed. Further researches for engineering research and design are also suggested

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

PSA in design of passive/active safety reactors

International Nuclear Information System (INIS)

Sato, T.; Tanabe, A.; Kondo, S.

1995-01-01

PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

Condition Based Prognostics of Passive Components - A New Era for Nuclear Power Plant Life Management

International Nuclear Information System (INIS)

Bakhtiari, S.; Mohanty, S.; Prokofiev, I.; Tregoning, R.

2012-01-01

As part of a research project sponsored by the U.S. NRC, Argonne National Laboratory (ANL) conducted scoping studies to identify viable and promising sensors and techniques for in-situ inspection and real-time monitoring of degradation in nuclear power plant (NPP) systems, structures, and components (SSC). Significant advances have been made over the past two decades toward development of online monitoring (OLM) techniques for detection, diagnostics, and prognostics of degradation in active nuclear power plant (NPP) components (e.g., pumps, valves). However, early detection of damage and degradation in safety-critical passive components, (e.g. piping, tubing pressure vessel), is challenging, and will likely remain so for the foreseeable future. Ensuring the structural integrity of the reactor pressure vessel (RPV) and piping systems in particular is a prerequisite to long term safe operation of NPPs. The current practice is to implement inservice inspection (ISI) and preventive maintenance programs. While these programs have generally been successful, they are limited in that information is only obtained during plant outages. Additionally, these inspections, often the critical path in the outage schedule, are costly, time consuming, and involve potentially high dose to nondestructive examination/evaluation (NDE) personnel. A viable plant-wide on-line structural health monitoring program for continuous and automatic monitoring of critical SSCs could be a more effective approach for guarding against unexpected failures. Specifically, OLM information about the current condition of the SSCs could be input to an online prognostics (OLP) system to forecast their remaining useful life in real time. This paper provides an overview of scoping studies performed at ANL on assessing the viability of OLM and OLP systems for real time and automated monitoring and remaining of condition and the remaining useful life of passive components in NPPs. (author)

The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR

International Nuclear Information System (INIS)

Sanchez J, J.; Morales S, J. B.

2009-10-01

A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

Invariant methods for an ensemble-based sensitivity analysis of a passive containment cooling system of an AP1000 nuclear power plant

International Nuclear Information System (INIS)

Di Maio, Francesco; Nicola, Giancarlo; Borgonovo, Emanuele; Zio, Enrico

2016-01-01

Sensitivity Analysis (SA) is performed to gain fundamental insights on a system behavior that is usually reproduced by a model and to identify the most relevant input variables whose variations affect the system model functional response. For the reliability analysis of passive safety systems of Nuclear Power Plants (NPPs), models are Best Estimate (BE) Thermal Hydraulic (TH) codes, that predict the system functional response in normal and accidental conditions and, in this paper, an ensemble of three alternative invariant SA methods is innovatively set up for a SA on the TH code input variables. The ensemble aggregates the input variables raking orders provided by Pearson correlation ratio, Delta method and Beta method. The capability of the ensemble is shown on a BE–TH code of the Passive Containment Cooling System (PCCS) of an Advanced Pressurized water reactor AP1000, during a Loss Of Coolant Accident (LOCA), whose output probability density function (pdf) is approximated by a Finite Mixture Model (FMM), on the basis of a limited number of simulations. - Highlights: • We perform the reliability analysis of a passive safety system of Nuclear Power Plant (NPP). • We use a Thermal Hydraulic (TH) code for predicting the NPP response to accidents. • We propose an ensemble of Invariant Methods for the sensitivity analysis of the TH code • The ensemble aggregates the rankings of Pearson correlation, Delta and Beta methods. • The approach is tested on a Passive Containment Cooling System of an AP1000 NPP.

Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

2013-09-01

This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

Enhanced Passive Cooling for Waterless-Power Production Technologies

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-06-14

Recent advances in the literature and at SNL indicate the strong potential for passive, specialized surfaces to significantly enhance power production output. Our exploratory computational and experimental research indicates that fractal and swirl surfaces can help enable waterless-power production by increasing the amount of heat transfer and turbulence, when compared with conventional surfaces. Small modular reactors, advanced reactors, and non-nuclear plants (e.g., solar and coal) are ideally suited for sCO2 coolant loops. The sCO2 loop converts the thermal heat into electricity, while the specialized surfaces passively and securely reject the waste process heat in an environmentally benign manner. The resultant, integrated energy systems are highly suitable for small grids, rural areas, and arid regions.

Passive decay heat removal from the core region

International Nuclear Information System (INIS)

Hichen, E.F.; Jaegers, H.

2002-01-01

The decay heat in commercial Light Water Reactors is commonly removed by active and redundant safety systems supported by emergency power. For advanced power plant designs passive safety systems using a natural circulation mode are proposed: several designs are discussed. New experimental data gained with the NOKO and PANDA facilities as well as operational data from the Dodewaard Nuclear Power Plant are presented and compared with new calculations by different codes. In summary, the effectiveness of these passive decay heat removal systems have been demonstrated: original geometries and materials and for the NOKO facility and the Dodewaard Reactor typical thermal-hydraulic inlet and boundary conditions have been used. With several codes a good agreement between calculations and experimental data was achieved. (author)

Design and analysis of new prestressed concrete containment and its passive cooling system for nuclear power plants

International Nuclear Information System (INIS)

Tan Xiaoshi; Li Xiaowei; Li Xiaotian; He Shuyan

2014-01-01

A new nuclear power plant prestressed concrete containment and its passive cooling system design were proposed for CAP1700 nuclear power plant as an example. The thermal-hydraulic calculation method for the new passive containment cooling system of CAP1700 was introduced and the operating parameters in accident condition were obtained. The result shows that the design of passive containment cooling system for CAP1700 is feasible and can meet the cooling demand in accident condition. Reservoir capacity of tank has a big margin and can be further optimized by calculation. (authors)

The role of passive and inherent safety properties in Siemens/KWU nuclear power plants

International Nuclear Information System (INIS)

Gremm, O.

1990-01-01

In Siemens/KWU Nuclear Power Plants the applied safety concept consist of a well balanced combination of active, passive use well is inherent safety measures. In principle it is not possible to realise a safety concept exclusively with inherent and/or passive safety properties. The respective measures and arguments will be explained in detail in the presentation. In addition the Siemens/KWU safety concept with examples of the role of inherent and passive safety measures will be illustrated. (author). 9 refs, 9 figs

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.

Reactivity control in HTR power plants with respect to passive safety system. Summary

Energy Technology Data Exchange (ETDEWEB)

Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

1996-12-01

The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).

New reactor programs from passive to pebble bed

Energy Technology Data Exchange (ETDEWEB)

Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

2002-07-01

The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

New reactor programs from passive to pebble bed

International Nuclear Information System (INIS)

Bruschi, H.J.

2002-01-01

The market for new nuclear power plants is small and challenged by alternative means of electric power generation. Customers and countries may vary in their requirements for a new nuclear plant; but all have a common theme of seeking a design that possesses favorable economics. This paper sets forth the economic challenges a new nuclear plant must overcome. In particular, it delineates the capital cost, construction time, and generation cost required to compete with combined cycle gas electric power generation. The U.S. power generation market is used as a point of comparison. Following this, the portfolio of BNFL/ Westinghouse plant designs are described and the methods by which they will meet the economic challenges previously delineated will be discussed. The portfolio includes the family of passive plants originated by the AP600 Design Certification process in the U.S. These plants are marked by a high degree of safety and simplicity, short construction times, and superior economics. In addition, the effort to meet European requirements for passive plants will be described. Lastly, the paper explores some advanced nuclear designs that are not yet licensed, and the hope that they hold for meeting the industry challenge ahead. (author)

Advanced passivation techniques for Si solar cells with high-κ dielectric materials

International Nuclear Information System (INIS)

Geng, Huijuan; Lin, Tingjui; Letha, Ayra Jagadhamma; Hwang, Huey-Liang; Kyznetsov, Fedor A.; Smirnova, Tamara P.; Saraev, Andrey A.; Kaichev, Vasily V.

2014-01-01

Electronic recombination losses at the wafer surface significantly reduce the efficiency of Si solar cells. Surface passivation using a suitable thin dielectric layer can minimize the recombination losses. Herein, advanced passivation using simple materials (Al 2 O 3 , HfO 2 ) and their compounds H (Hf) A (Al) O deposited by atomic layer deposition (ALD) was investigated. The chemical composition of Hf and Al oxide films were determined by X-ray photoelectron spectroscopy (XPS). The XPS depth profiles exhibit continuous uniform dense layers. The ALD-Al 2 O 3 film has been found to provide negative fixed charge (−6.4 × 10 11  cm −2 ), whereas HfO 2 film provides positive fixed charge (3.2 × 10 12  cm −2 ). The effective lifetimes can be improved after oxygen gas annealing for 1 min. I-V characteristics of Si solar cells with high-κ dielectric materials as passivation layers indicate that the performance is significantly improved, and ALD-HfO 2 film would provide better passivation properties than that of the ALD-Al 2 O 3 film in this research work.

Thermal-hydraulic modeling needs for passive reactors

Energy Technology Data Exchange (ETDEWEB)

Kelly, J.M. [Nuclear Regulatory Commission, Washington, DC (United States)

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.

Thermal-hydraulic modeling needs for passive reactors

International Nuclear Information System (INIS)

Kelly, J.M.

1997-01-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered, but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken

Pressure suppression pool mixing in passive advanced BWR plants

International Nuclear Information System (INIS)

Gamble, Robert E.; Nguyen, Thuy T.; Shiralkar, Bharat S.; Peterson, Per F.; Greif, Ralph; Tabata, H.

2001-01-01

In the SBWR passive boiling water reactor, the long-term post-accident containment pressure is determined by the combination of noncondensible gas pressure and steam pressure in the wetwell gas space. The suppression pool (SP) surface temperature, which determines the vapor partial pressure, is very important to overall containment performance. Therefore, the thermal stratification of the SP due to blowdown is of primary importance. This work looks at the various phases and phenomena present during the blowdown event and identifies those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests. This is important in determining which of the large body of blowdown to SP data is adequate for application to the stratification problem. The mixing by jets from the main vents is identified as the key phenomena influencing the thermal response of the suppression pool and analytical models are developed to predict the jet influence on thermal stratification. The analytical models are implemented into a system simulation code, TRACG, and used to model thermal stratification behavior in a scaled test facility. The results show good general agreement with the test data

Advancing CANDU Technology Through R and D

International Nuclear Information System (INIS)

Torgerson, David F.

1993-01-01

CANDU reactors are evolving to meet future requirements using incremental changes as opposed to revolutionary design changes. The main elements for advancing the technology reducing capital and operating, increasing capacity factors, increasing passive safety, and enhancing fuel/fuel cycle flexibility. These elements are being addressed by carrying out research and development in the areas of safety, plant systems and components, heavy water production, information technology, fuel channels, and fuel/fuel cycle technology. In safety, the focus is on using the inherent features of CANDU to enhance passive or natural safety concepts, such as the use of the moderator as an effective heat sink, and the development of advanced fuels to improve critical heat flux and to reduce source terms. Plant systems and components work includes improvements to plant systems such as steam generators, heat exchangers, pump seals, and advanced control room technology. Heavy water processes are being developed that can be used with existing hydrogen production plants, or that can be used in a stand-alone mode. Information technology is being developed to cover all aspects of CANDU design, construction, and operation. Fuel channel improvements include elucidation and application of basic materials science for life extension, and the development of advanced non-destructive examination methods. Fuel and fuel cycle work is focusing on LWR/CANDU synergy, such as the use of recovered uranium and the direct use of spent PWR fuel in CANDU reactor, advanced fuels to improve burnup and economics (e. g., the joint AECB/KAERI Conflux program), and low void reactivity fuel to enhance passive safety. This paper gives an overview of some of the R and D supporting these activities, with particular emphasis on Alice's vision for advancing CANDU technology over the next 10 years

Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

International Nuclear Information System (INIS)

Bruschi, H.J.

1996-01-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

Energy Technology Data Exchange (ETDEWEB)

Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1996-10-01

At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

International Nuclear Information System (INIS)

Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

2015-01-01

As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems

International Nuclear Information System (INIS)

Carsten, Pasler

2007-01-01

The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand

Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin

2015-06-28

Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

U. S. Utility Leadership in Requirements For Passive Reactors

International Nuclear Information System (INIS)

Kim, Jcng H.; Layman, William H.

1991-01-01

Utility leadership from both U.S. utilities and international utilities, is a key element in the U. S. Advanced Light Water Reactor Program. International utilities have played a very import Design reviews by the utilities participating in the ALRR Program will ensure that all of the utility requirements are met while design work is being carried out. Our mission is to achieve NRC certification of designs that reflect the needs of the utilities and we believe that this will play an important role in the resurgence of nuclear plant construction in the United States. As stated in the Nuclear Power Oversight Committee's Strategic Plan For Building New Nuclear Power Plants : 'The extensive operating experience with today's light water reactors (LWRs), and the promise shown in recent technical developments, leads the industry to the conclusion that the next nuclear plants ordered in the United States will be advanced light water reactors (A LWRs). Two types are under development : units of large output (1300 MW) called 'evolutionary' A LWRs and units of mid-size output (600 MW) called 'Passive' A LWRs. The term 'passive' refers to the safety features which depend more on natural processes such as gravity and buoyancy than on powered equipment such as pumps

Oregon state university's advanced plant experiment (APEX) AP1000 integral facility test program

International Nuclear Information System (INIS)

Reyes, J.N.; Groome, J.T.; Woods, B.G.; Young, E.; Abel, K.; Wu, Q.

2005-01-01

Oregon State University (OSU) has recently completed a three year study of the thermal hydraulic behavior of the Westinghouse AP1000 passive safety systems. Eleven Design Basis Accident (DBA) scenarios, sponsored by the U.S. Department of Energy (DOE) with technical support from Westinghouse Electric, were simulated in OSU's Advanced Plant Experiment (APEX)-1000. The OSU test program was conducted within the purview of the requirements of 10CFR50 Appendix B, NQA-1 and 10 CFR 21 and the test data was used to provide benchmarks for computer codes used in the final design approval of the AP1000. In addition to the DOE certification testing, OSU conducted eleven confirmatory tests for the U.S. Nuclear Regulatory Commission. This paper presents the test program objectives, a description of the APEX-1000 test facility and an overview of the test matrix that was conducted in support of plant certification. (authors)

Advanced liquid metal reactor plant control system

International Nuclear Information System (INIS)

Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

1993-01-01

The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

International Nuclear Information System (INIS)

Schene, R.

2009-01-01

Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

Advanced plant design recommendations from Cook Nuclear Plant experience

International Nuclear Information System (INIS)

Zimmerman, W.L.

1993-01-01

A project in the American Electric Power Service Corporation to review operating and maintenance experience at Cook Nuclear Plant to identify recommendations for advanced nuclear plant design is described. Recommendations so gathered in the areas of plant fluid systems, instrument and control, testing and surveillance provisions, plant layout of equipment, provisions to enhance effective maintenance, ventilation systems, radiological protection, and construction, are presented accordingly. An example for a design review checklist for effective plant operations and maintenance is suggested

Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

International Nuclear Information System (INIS)

Forsberg, C.W.; Reich, W.J.

1991-09-01

The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

Simplified nuclear plant design for tomorrow's energy needs

International Nuclear Information System (INIS)

Slember, R.

1989-09-01

Commercial nuclear powered plants play an important role in the strategic energy plans of many countries throughout the world. Many energy planners agree that nuclear plants will have to supply an increasing amount of electrical energy in the 1990s and beyond. Just as other major industries are continually taking steps to update and improve existing products, the United States' nuclear industry has embarked on a program to simplify plant systems, shorten construction time and improve economics for new plant models. One of the models being developed by Westinghouse Electric Corporation and Burns and Roe Company is the Advanced Passive 600 MWe design which incorporates safety features that passively protect the reactor during assumed abnormal operating events. These passive safety systems utilize natural circulation/cooling for mitigating abnormal events and simplify plant design and operation. This type of system eliminates the need for costly active safety grade components, results in a reduction of ancillary equipment and assists in shortening construction time. The use of passive safety systems also permits design simplification of the auxiliary systems effectively reducing operating and maintenance requirements. Collectively, the AP600 design features result in a safe plant that addresses and alleviates the critical industry issues that developed in the 1980s. Further, the design addresses utility and regulatory requirements for safety, reliability, maintainability, operations and economics. Program results to date give confidence that the objectives of the Advanced Passive 600 design are achievable through overall plant simplification. The report will include timely results from the work being performed on the salient technical features of the design, plant construction and operation. Other required institutional changes, such as the prerequisite for a design which is complete and licensed prior to start of construction, will also be presented

Inherent/passive safety in fusion power plants

International Nuclear Information System (INIS)

Piet, S.J.; Crocker, J.G.

1986-01-01

The concept of inherent or passive safety for fusion energy is explored, defined, and partially quantified. Four levels of safety assurance are defined, which range from true inherent safety to passive safety to protection via active engineered safeguard systems. Fusion has the clear potential for achieving inherent or passive safety, which should be an objective of fusion research and design. Proper material choice might lead to both inherent/passive safety and high mass power density, improving both safety and economics. When inherent or passive safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be a truly superior energy source for the future

Use of color-change indicators to quantify passive films on the stainless steel valves of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Cheng, Cong Qian [School of Materials Science, Engineering, Dalian University of Technology, Dalian 116085 (China); Yang, Shu Kai [School of Materials Science, Engineering, Dalian University of Technology, Dalian 116085 (China); Suzhou Nuclear Power Research Institute, Suzhou 215004 (China); Zhao, Jie, E-mail: jiezhao@dlut.edu.cn [School of Materials Science, Engineering, Dalian University of Technology, Dalian 116085 (China)

2016-02-15

Highlights: • A facile method to evaluate passivation quality by color change indicator. • Two indicators were compared in lab and applied on vales in nuclear power plants. • It shows that the higher value of color change the worse quality of passivation. • Traditional ferroxyl solution is unstable and might impair the vale surface. • The new indicator is more practicable than the ferroxyl test for on-site inspection. - Abstract: The passive film on nuclear-grade stainless steels was evaluated by quantifying its color changes. Coloration reactions were compared by using ferroin and blue dot solutions as indicators on the basis of the measured results in a laboratory. The reactions were then applied on stainless steel valves in a nuclear power plant. The degree of color change indicates the degree of growth of a passive film. The ferroin solution exhibits higher accuracy and more stable than blue dot solution in determining passive film quality. The potentiodynamic polarization curves show that blue dot solution might cause surface damage compared with ferroin solution. The inspection result on stainless steel valves supports our laboratory result. However, stainless steel exhibited a dramatic decrease in sensitivity to blue dot because of the intrinsic instability and high acidity of this solution. Ferroin solution is superior to blue dot solution for stainless steel facilities in a nuclear power plant.

The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

International Nuclear Information System (INIS)

Demetri, K. J.; Leipner, C. I.; Marshall, M. L.

2015-09-01

The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

Energy Technology Data Exchange (ETDEWEB)

Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2015-09-15

The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

Approaches to passive safety in advanced thermal reactors

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

Some Recent Advances in Plant Physiology

Science.gov (United States)

Stafford, G. A.

1972-01-01

A popular review of plant physiological research, emphasizing those apsects of plant metabolism where there has been a recent shift in emphasis that is not yet reflected in secondary school advanced texts. (AL)

Advancing the Use of Passive Sampling in Risk Assessment and Management of Sediments Contaminated with Hydrophobic Organic Chemicals: Results of an International Ex Situ Passive Sampling Interlaboratory Comparison.

Science.gov (United States)

Jonker, Michiel T O; van der Heijden, Stephan A; Adelman, Dave; Apell, Jennifer N; Burgess, Robert M; Choi, Yongju; Fernandez, Loretta A; Flavetta, Geanna M; Ghosh, Upal; Gschwend, Philip M; Hale, Sarah E; Jalalizadeh, Mehregan; Khairy, Mohammed; Lampi, Mark A; Lao, Wenjian; Lohmann, Rainer; Lydy, Michael J; Maruya, Keith A; Nutile, Samuel A; Oen, Amy M P; Rakowska, Magdalena I; Reible, Danny; Rusina, Tatsiana P; Smedes, Foppe; Wu, Yanwen

2018-03-20

This work presents the results of an international interlaboratory comparison on ex situ passive sampling in sediments. The main objectives were to map the state of the science in passively sampling sediments, identify sources of variability, provide recommendations and practical guidance for standardized passive sampling, and advance the use of passive sampling in regulatory decision making by increasing confidence in the use of the technique. The study was performed by a consortium of 11 laboratories and included experiments with 14 passive sampling formats on 3 sediments for 25 target chemicals (PAHs and PCBs). The resulting overall interlaboratory variability was large (a factor of ∼10), but standardization of methods halved this variability. The remaining variability was primarily due to factors not related to passive sampling itself, i.e., sediment heterogeneity and analytical chemistry. Excluding the latter source of variability, by performing all analyses in one laboratory, showed that passive sampling results can have a high precision and a very low intermethod variability (sampling, irrespective of the specific method used, is fit for implementation in risk assessment and management of contaminated sediments, provided that method setup and performance, as well as chemical analyses are quality-controlled.

Sodium hydroxide injection passivation work for the reactor water clean-up system in a new ABWR plant

International Nuclear Information System (INIS)

Wen, Tung-Jen; Lu, Ju-Huang

2012-09-01

Several studies have identified that Co-58 and Co-60 as the primary source of radiation build up on out-of-core components in new BWR plants. The deposition rate of Co on stainless steel and carbon steel is shown to be controlled mainly by the thickness of oxide films and its morphology formed through pretreatment. The passivation treatment was implemented accordingly at Lungmen unit 1 in an ABWR plant in September 2010. It is determined that the passivation conditions should be maintained at the temperature of 180∼230 deg. C, pH of 8.0∼8.5 and dissolved oxygen content over 400 ppb. The films would provide effective protection against radioactive deposition. The application of the pre-filming process on piping before the pre-operation is done during the flow induced vibration test (FIV) period. The protectiveness of stable magnetite can be increased by the pH control under the specific condition. The pre-filming control process and evaluation of passivation effectiveness is discussed in detail based on the surface analysis of the passivated specimens. Many efforts have been devoted to sodium hydroxide injection method for pH control of the system through the filter demineralizer under smooth operation. A comparison of test specimens on the properties of oxide film formed between laboratory and in-plant tests through alkaline treatment are also shown in this report. (authors)

The research activities on in-tube condensation in the presence of noncondensables for passive cooling applications

Energy Technology Data Exchange (ETDEWEB)

Tanrikut, A [Turkish Atomic Energy Authority, Ankara (Turkey)

1996-12-01

The introduction of nuclear power becomes an attractive solution to the problem of increasing demand for electricity power capacity in Turkey. Thus, Turkey is willing to follow the technological development trends in advanced reactor systems and to participate in joint research studies. The primary objectives of the passive design features are to simplify the design, which assures the minimized demand on operator, and to improve plant safety. To accomplish these features the operating principles of passive safety systems should be well understood by an experimental validation program. Such a validation program is also important for the assessment of advanced computer codes which are currently used for design and licensing procedures. The condensation mode of heat transfer plays an important role for the passive heat removal applications in the current nuclear power plants (e.g. decay heat removal via steam generators in case of loss of heat removal system) and advanced water-cooled reactor systems. But is well established that the presence of noncondensable gases can greatly inhibit the condensation process due to the build-up of noncondensable gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of noncondensable. This paper describes the research activities at the Turkish Atomic Energy Authority concerning condensation in the presence of air, as a noncondensable gas. (author). 9 refs, 6 figs.

The research activities on in-tube condensation in the presence of noncondensables for passive cooling applications

International Nuclear Information System (INIS)

Tanrikut, A.

1996-01-01

The introduction of nuclear power becomes an attractive solution to the problem of increasing demand for electricity power capacity in Turkey. Thus, Turkey is willing to follow the technological development trends in advanced reactor systems and to participate in joint research studies. The primary objectives of the passive design features are to simplify the design, which assures the minimized demand on operator, and to improve plant safety. To accomplish these features the operating principles of passive safety systems should be well understood by an experimental validation program. Such a validation program is also important for the assessment of advanced computer codes which are currently used for design and licensing procedures. The condensation mode of heat transfer plays an important role for the passive heat removal applications in the current nuclear power plants (e.g. decay heat removal via steam generators in case of loss of heat removal system) and advanced water-cooled reactor systems. But is well established that the presence of noncondensable gases can greatly inhibit the condensation process due to the build-up of noncondensable gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of noncondensable. This paper describes the research activities at the Turkish Atomic Energy Authority concerning condensation in the presence of air, as a noncondensable gas. (author). 9 refs, 6 figs

Comparison of advanced mid-sized reactors regarding passive features, core damage frequencies and core melt retention features

International Nuclear Information System (INIS)

Wider, H.

2005-01-01

New Light Water Reactors, whose regular safety systems are complemented by passive safety systems, are ready for the market. The special aspect of passive safety features is their actuation and functioning independent of the operator. They add significantly to reduce the core damage frequency (CDF) since the operator continues to play its independent role in actuating the regular safety devices based on modern instrumentation and control (I and C). The latter also has passive features regarding the prevention of accidents. Two reactors with significant passive features that are presently offered on the market are the AP1000 PWR and the SWR 1000 BWR. Their passive features are compared and also their core damage frequencies (CDF). The latter are also compared with those of a VVER-1000. A further discussion about the two passive plants concerns their mitigating features for severe accidents. Regarding core-melt retention both rely on in-vessel cooling of the melt. The new VVER-1000 reactor, on the other hand features a validated ex-vessel concept. (author)

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)

Predicting plant attractiveness to pollinators with passive crowdsourcing.

Science.gov (United States)

Bahlai, Christie A; Landis, Douglas A

2016-06-01

observations were not associated with Internet images, but were slightly associated with BP. Our results suggest that passively crowdsourced image data can potentially be a useful screening tool to identify candidate plants for pollinator habitat restoration efforts directed at wild bee conservation. Increasing our understanding of the attractiveness of a greater diversity of plants increases the potential for more rapid and efficient research in creating pollinator-supportive landscapes.

Advanced Performance Modeling with Combined Passive and Active Monitoring

Energy Technology Data Exchange (ETDEWEB)

Dovrolis, Constantine [Georgia Inst. of Technology, Atlanta, GA (United States); Sim, Alex [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2015-04-15

To improve the efficiency of resource utilization and scheduling of scientific data transfers on high-speed networks, the "Advanced Performance Modeling with combined passive and active monitoring" (APM) project investigates and models a general-purpose, reusable and expandable network performance estimation framework. The predictive estimation model and the framework will be helpful in optimizing the performance and utilization of networks as well as sharing resources with predictable performance for scientific collaborations, especially in data intensive applications. Our prediction model utilizes historical network performance information from various network activity logs as well as live streaming measurements from network peering devices. Historical network performance information is used without putting extra load on the resources by active measurement collection. Performance measurements collected by active probing is used judiciously for improving the accuracy of predictions.

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

Applied reliability assessment for the passive safety systems of nuclear power plants (NPPs) using system dynamics (SD)

International Nuclear Information System (INIS)

Kim, Yun Il; Woo, Tae Ho

2018-01-01

The passive system by the free-fall is investigated in the accident of nuclear power plants (NPPs). The complex algorithm of the system dynamics (SD) modeling is done in the passive cooling system. The nuclear passive system by free-fall is successfully modeled for the loss of coolant accident (LOCA). Conventional passive system of gravity or natural circulation is working only when the piping systems is in the good condition. The external coolant supply system is introduced in the case of the piping system failure. The water is poured into the reactor through the guiding piping or tube. If the explosion happens, the coolants could be showering into the reactor core and its building. New kind of passive system is expected successfully in the on-site black out where the drone could be operated by battery or engine.

Technical feasibility and reliability of passive safety systems of AC600

International Nuclear Information System (INIS)

Niu, W.; Zeng, X.

1996-01-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished by the Nuclear Power Institute of China. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also

Technical feasibility and reliability of passive safety systems of AC600

Energy Technology Data Exchange (ETDEWEB)

Niu, W; Zeng, X [Nuclear Power Inst. of China, Chendu (China)

1996-12-01

The first step conceptual design of the 600 MWe advanced PWR (AC-600) has been finished. Experiments on the passive system of AC-600 are being carried out, and are expected to be completed next year. The main research emphases of AC-600 conceptual design include the advanced core, the passive safety system and simplification. The design objective of AC-600 is that the safety, reliability, maintainability, operation cost and construction period are all improved upon compared to those of PWR plant. One of important means to achieve the objective is using a passive system, which has the following functions whenever its operation is required: providing the reactor core with enough coolant when others fail to make up the lost coolant; reactor residual heat removal; cooling and reducing pressure in the containment and preventing radioactive substances from being released into the environment after occurrence of accident (e.g. LOCA). The system should meet the single failure criterion, and keep operating when a single active component or passive component breaks down during the first 72 hour period after occurrence of accident, or in the long period following the 72 hour period. The passive safety system of AC-600 is composed of the primary safety injection system, the secondary emergency core residual heat removal system and the containment cooling system. The design of the system follows some relevant rules and criteria used by current PWR plant. The system has the ability to bear single failure, two complete separate subsystems are considered, each designed for 100% working capacity. Normal operation is separate from safety operation and avoids cross coupling and interference between systems, improves the reliability of components, and makes it easy to maintain, inspect and test the system. The paper discusses the technical feasibility and reliability of the passive safety system of AC-600, and some issues and test plans are also involved. (author). 3 figs, 1 tab.

Materials for passively safe reactors

International Nuclear Information System (INIS)

Simnad, T.

1993-01-01

Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

Passive Safety Features for Small Modular Reactors

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2010-01-01

The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accidents consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

Concept research on general passive system

International Nuclear Information System (INIS)

Han Xu; Yang Yanhua; Zheng Mingguang

2009-01-01

This paper summarized the current passive techniques used in nuclear power plants. Through classification and analysis, the functional characteristics and inherent identification of passive systems were elucidated. By improving and extending the concept of passive system, the general passive concept was proposed, and space and time relativity was discussed and assumption of general passive system were illustrated. The function of idealized general passive system is equivalent with the current passive system, but the design of idealized general passive system is more flexible. (authors)

Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06

International Nuclear Information System (INIS)

2006-01-01

Following the highly successful ICAPP'05 meeting held in Seoul Korea, the 2006 International Congress on Advances in Nuclear Power Plants brought together international experts of the nuclear industry involved in the operation, development, building, regulation and research related to Nuclear Power Plants. The program covers the full spectrum of Nuclear Power Plant issues from design, deployment and construction of plants to research and development of future designs and advanced systems. The program covers lessons learned from power, research and demonstration reactors from over 50 years of experience with operation and maintenance, structures, materials, technical specifications, human factors, system design and reliability. The program by technical track deals with: - 1. Water-Cooled Reactor Programs and Issues Evolutionary designs, innovative, passive, light and heavy water cooled reactors; issues related to meeting medium term utility needs; design and regulatory issues; business, political and economic challenges; infrastructure limitations and improved construction techniques including modularization. - 2. High Temperature Gas Cooled Reactors Design and development issues, components and materials, safety, reliability, economics, demonstration plants and environmental issues, fuel design and reliability, power conversion technology, hydrogen production and other industrial uses; advanced thermal and fast reactors. - 3. Long Term Reactor Programs and Strategies Reactor technology with enhanced fuel cycle features for improved resource utilization, waste characteristics, and power conversion capabilities. Potential reactor designs with longer development times such as, super critical water reactors, liquid metal reactors, gaseous and liquid fuel reactors, Gen IV, INPRO, EUR and other programs. - 4. Operation, Performance and Reliability Management Training, O and M costs, life cycle management, risk based maintenance, operational experiences, performance and

Operating experiences with passive systems and components in German nuclear power plants

International Nuclear Information System (INIS)

Maqua, M.

1996-01-01

Operating experience with passive systems and components is limited to the equipment installed in existing NPPs. In German power plants, this experience is available for equipment of the IAEA categories A, C and D. The presentation is focused on typical examples out of these three categories. An overview is given on the number of reported events and typical failure modes. Selected failures are discussed in detail. 1 ref., 6 figs, 7 tabs

Operating experiences with passive systems and components in German nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Maqua, M [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

1996-12-01

Operating experience with passive systems and components is limited to the equipment installed in existing NPPs. In German power plants, this experience is available for equipment of the IAEA categories A, C and D. The presentation is focused on typical examples out of these three categories. An overview is given on the number of reported events and typical failure modes. Selected failures are discussed in detail. 1 ref., 6 figs, 7 tabs.

Fate of As, Se, and Hg in a Passive Integrated System for Treatment of Fossil Plant Wastewater

Energy Technology Data Exchange (ETDEWEB)

Terry Yost; Paul Pier; Gregory Brodie

2007-12-31

TVA is collaborating with EPRI and DOE to demonstrate a passive treatment system for removing SCR-derived ammonia and trace elements from a coal-fired power plant wastewater stream. The components of the integrated system consist of trickling filters for ammonia oxidation, reaction cells containing zero-valent iron (ZVI) for trace contaminant removal, a settling basin for storage of iron hydroxide floc, and anaerobic vertical-flow wetlands for biological denitrification. The passive integrated treatment system will treat up to 0.25 million gallons per day (gpd) of flue gas desulfurization (FGD) pond effluent, with a configuration requiring only gravity flow to obviate the need for pumps. The design of the system will enable a comparative evaluation of two parallel treatment trains, with and without the ZVI extraction trench and settling/oxidation basin components. One of the main objectives is to gain a better understanding of the chemical transformations that species of trace elements such as arsenic, selenium, and mercury undergo as they are treated in passive treatment system components with differing environmental conditions. This progress report details the design criteria for the passive integrated system for treating fossil power plant wastewater as well as performance results from the first several months of operation. Engineering work on the project has been completed, and construction took place during the summer of 2005. Monitoring of the passive treatment system was initiated in October 2005 and continued until May 18 2006. The results to date indicate that the treatment system is effective in reducing levels of nitrogen compounds and trace metals. Concentrations of both ammonia and trace metals were lower than expected in the influent FGD water, and additions to increase these concentrations will be done in the future to further test the removal efficiency of the treatment system. In May 2006, the wetland cells were drained of FGD water, refilled with

Passive residual energy utilization system in thermal cycles on water-cooled power reactors

International Nuclear Information System (INIS)

Placco, Guilherme M.; Guimaraes, Lamartine N.F.; Santos, Rubens S. dos

2013-01-01

This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

Advanced Power Plant Development and Analyses Methodologies

Energy Technology Data Exchange (ETDEWEB)

G.S. Samuelsen; A.D. Rao

2006-02-06

Under the sponsorship of the U.S. Department of Energy/National Energy Technology Laboratory, a multi-disciplinary team led by the Advanced Power and Energy Program of the University of California at Irvine is defining the system engineering issues associated with the integration of key components and subsystems into advanced power plant systems with goals of achieving high efficiency and minimized environmental impact while using fossil fuels. These power plant concepts include ''Zero Emission'' power plants and the ''FutureGen'' H{sub 2} co-production facilities. The study is broken down into three phases. Phase 1 of this study consisted of utilizing advanced technologies that are expected to be available in the ''Vision 21'' time frame such as mega scale fuel cell based hybrids. Phase 2 includes current state-of-the-art technologies and those expected to be deployed in the nearer term such as advanced gas turbines and high temperature membranes for separating gas species and advanced gasifier concepts. Phase 3 includes identification of gas turbine based cycles and engine configurations suitable to coal-based gasification applications and the conceptualization of the balance of plant technology, heat integration, and the bottoming cycle for analysis in a future study. Also included in Phase 3 is the task of acquiring/providing turbo-machinery in order to gather turbo-charger performance data that may be used to verify simulation models as well as establishing system design constraints. The results of these various investigations will serve as a guide for the U. S. Department of Energy in identifying the research areas and technologies that warrant further support.

Advanced Power Plant Development and Analysis Methodologies

Energy Technology Data Exchange (ETDEWEB)

A.D. Rao; G.S. Samuelsen; F.L. Robson; B. Washom; S.G. Berenyi

2006-06-30

Under the sponsorship of the U.S. Department of Energy/National Energy Technology Laboratory, a multi-disciplinary team led by the Advanced Power and Energy Program of the University of California at Irvine is defining the system engineering issues associated with the integration of key components and subsystems into advanced power plant systems with goals of achieving high efficiency and minimized environmental impact while using fossil fuels. These power plant concepts include 'Zero Emission' power plants and the 'FutureGen' H2 co-production facilities. The study is broken down into three phases. Phase 1 of this study consisted of utilizing advanced technologies that are expected to be available in the 'Vision 21' time frame such as mega scale fuel cell based hybrids. Phase 2 includes current state-of-the-art technologies and those expected to be deployed in the nearer term such as advanced gas turbines and high temperature membranes for separating gas species and advanced gasifier concepts. Phase 3 includes identification of gas turbine based cycles and engine configurations suitable to coal-based gasification applications and the conceptualization of the balance of plant technology, heat integration, and the bottoming cycle for analysis in a future study. Also included in Phase 3 is the task of acquiring/providing turbo-machinery in order to gather turbo-charger performance data that may be used to verify simulation models as well as establishing system design constraints. The results of these various investigations will serve as a guide for the U. S. Department of Energy in identifying the research areas and technologies that warrant further support.

ARIES-AT: An advanced tokamak, advanced technology fusion power plant

International Nuclear Information System (INIS)

Najmabadi, F.; Jardin, S.C.; Tillack, M.; Waganer, L.M.

2001-01-01

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant. Several avenues were pursued in order to arrive at plasmas with a higher β and better bootstrap alignment compared to ARIES-RS that led to plasmas with higher β N and β. Advanced technologies that are examined in detail include: (1) Possible improvements to the overall system by using high-temperature superconductors, (2) Innovative SiC blankets that lead to a high thermal cycle efficiency of ∼60%; and (3) Advanced manufacturing techniques which aim at producing near-finished products directly from raw material, resulting in low-cost, and reliable components. The 1000-MWe ARIES-AT design has a major radius of 5.4 m, minor radius of 1.3 M, a toroidal β of 9.2% (β N =6.0) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current drive power is 24 MW. The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (5c/kWh), which is competitive with those projected for other sources of energy. (author)

Proceedings of the 2008 International Congress on Advances in Nuclear Power Plants - ICAPP '08

International Nuclear Information System (INIS)

2008-01-01

ICAPP 2008 congress brought together international experts of the nuclear industry involved in the operation, development, building, regulation and research related to Nuclear Power Plants. The program covered the full spectrum of Nuclear Power Plant issues from design, deployment and construction of plants to research and development of future designs and advanced systems. It covered also lessons learned from power, research and demonstration reactors from over 50 years of experience with operation and maintenance, structures, materials, technical specifications, human factors, system design and reliability. The program comprised 13 technical tracks: 1. Water-Cooled Reactor Programs and Issues: Evolutionary designs, innovative, passive, light and heavy water cooled reactors; issues related to meeting near term utility needs; design issues; business, economical cost challenges; infrastructure limitations and improved construction techniques including modularization. 2. High Temperature Gas Cooled Reactors: Design and development issues, components and materials, safety, reliability, economics, demonstration plants and environmental issues, fuel design and reliability, power conversion technology, impact of non electricity applications on reactor design; advanced thermal and fast reactors. 3. LMFR and Longer Term Reactor Programs: Reactor technology with enhanced fuel cycle features for improved resource utilization, waste characteristics, and power conversion capabilities. Potential reactor designs with longer development times such as super critical water reactors and liquid fuel reactors, Gen IV, INPRO, EUR and other programs. 4. Operation, Performance and Reliability Management: Training, O and M costs, life cycle management, risk based maintenance, operational experiences, performance and reliability improvements, outage optimization, human factors, plant staffing, outage reduction features, major component reliability, repair and replacement, in

The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

Energy Technology Data Exchange (ETDEWEB)

Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

2009-10-15

A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

Passive safety systems reliability and integration of these systems in nuclear power plant PSA

International Nuclear Information System (INIS)

La Lumia, V.; Mercier, S.; Marques, M.; Pignatel, J.F.

2004-01-01

Innovative nuclear reactor concepts could lead to use passive safety features in combination with active safety systems. A passive system does not need active component, external energy, signal or human interaction to operate. These are attractive advantages for safety nuclear plant improvements and economic competitiveness. But specific reliability problems, linked to physical phenomena, can conduct to stop the physical process. In this context, the European Commission (EC) starts the RMPS (Reliability Methods for Passive Safety functions) program. In this RMPS program, a quantitative reliability evaluation of the RP2 system (Residual Passive heat Removal system on the Primary circuit) has been realised, and the results introduced in a simplified PSA (Probabilistic Safety Assessment). The scope is to get out experience of definition of characteristic parameters for reliability evaluation and PSA including passive systems. The simplified PSA, using event tree method, is carried out for the total loss of power supplies initiating event leading to a severe core damage. Are taken into account: failures of components but also failures of the physical process involved (e.g. natural convection) by a specific method. The physical process failure probabilities are assessed through uncertainty analyses based on supposed probability density functions for the characteristic parameters of the RP2 system. The probabilities are calculated by MONTE CARLO simulation coupled to the CATHARE thermalhydraulic code. The yearly frequency of the severe core damage is evaluated for each accident sequence. This analysis has identified the influence of the passive system RP2 and propose a re-dimensioning of the RP2 system in order to satisfy the safety probabilistic objectives for reactor core severe damage. (authors)

Development of passive condensers for accident localization systems at nuclear power plants in the former USSR

International Nuclear Information System (INIS)

Kuznecov, M.V.

1992-01-01

The development is summarized of passive condensers for accident localization systems at nuclear power plants (with RBMK and WWER reactors) in the former USSR. Basic properties and criteria defining their availability are described, as are experimental tests and technical solution optimization results. (author) 2 fig

Passive heat transport in advanced CANDU containment

International Nuclear Information System (INIS)

Krause, M.; Mathew, P.M.

1994-01-01

A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

An integral reactor design concept for a nuclear co-generation plant

International Nuclear Information System (INIS)

Lee, D.J.; Kim, J.I.; Kim, K.K.; Chang, M.H.; Moon, K.S.

1997-01-01

An integral reactor concept for nuclear cogeneration plant is being developed at KAERI as an attempt to expand the peaceful utilization of well established commercial nuclear technology, and related industrial infrastructure such as desalination technology in Korea. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway to evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design. This paper describes the preliminary safety and design concepts of the advanced integral reactor. Salient features of the design are hexagonal core geometry, once-through helical steam generator, self-pressurizer, and seismic resistant fine control CEDMS, passive residual heat removal system, steam injector driven passive containment cooling system. (author)

Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

International Nuclear Information System (INIS)

White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.

1989-01-01

Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs

New Frontiers in Passive and Active Nanoantennas

DEFF Research Database (Denmark)

Arslanagic, Samel; Ziolkowski, Richard W.

2017-01-01

The articles included in this special section focus on several recent advances in the field of passive and active nanoantennas that employ not only traditional based realizations but also their new frontiers.......The articles included in this special section focus on several recent advances in the field of passive and active nanoantennas that employ not only traditional based realizations but also their new frontiers....

SBO simulations for Integrated Passive Safety System (IPSS) using MARS

International Nuclear Information System (INIS)

Kim, Sang Ho; Jeong, Sung Yeop; Chang, Soon Heung

2012-01-01

The current nuclear power plants have lots of active safety systems with some passive safety systems. The safety of current and future nuclear power plants can be enhanced by the application of additional passive safety systems for the ultimate safety. It is helpful to install the passive safety systems on current nuclear power plants without the design change for the licensibility. For solving the problem about the system complexity shown in the Fukushima accidents, the current nuclear power plants are needed to be enhanced by an additional integrated and simplified system. As a previous research, the integrated passive safety system (IPSS) was proposed to solve the safety issues related with the decay heat removal, containment integrity and radiation release. It could be operated by natural phenomena like gravity, natural circulation and pressure difference without AC power. The five main functions of IPSS are: (a) Passive decay heat removal, (b) Passive emergency core cooling, (c) Passive containment cooling, (d) Passive in vessel retention and ex-vessel cooling, and (e) Filtered venting and pressure control. The purpose of this research is to analyze the performances of each function by using MARS code. The simulated accident scenarios were station black out (SBO) and the additional accidents accompanied by SBO

Advanced chemistry management system for nuclear power plants

International Nuclear Information System (INIS)

Maeda, Katsuji; Kobayashi, Yasuhiro; Nagasawa, Katsumi

2000-01-01

Chemistry control in a boiling water reactor (BWR) plant has a close relationship with radiation field buildup, fuel reliability, integrity of plant components and materials, performance of the water treatment systems and radioactive waste generation. Chemistry management in BWR plants has become more important in order to maintain and enhance plant reliability. Adequate chemistry control and management are also essential to establish, maintain, and enhance plant availability. For these reasons, we have developed the advanced chemistry management system for nuclear power plants in order to effectively collect and evaluate a large number of plant operating and chemistry data. (author)

Advanced plant engineering and construction of Japanese ABWRs

International Nuclear Information System (INIS)

Gotoh, N.; Sumikawa, J.; Yoshida, N.; Yoshida, M.

1998-01-01

Remarkable improvement has been made in recent nuclear power plant design and construction in Japan. These many improved engineering technologies has been made a good use in the lately commercial operated two world's first 1,356MWe ABW's (Advanced Boiling Water Reactors), and made a great contribution to the smooth progress and the completion of a highly reliable plant construction. Especially, two engineering technologies, (1), three-dimensional computer aided design system through engineering data-base, and (2), large scale modularising construction method, have been successfully applied as the integrated engineering technologies of the plant construction. And two integrated reviews, 'integrated design review, confirmation of new and changed design and prevention of failure recurrence' in the design stage, and 'constructing plant review' at the site, have been widely and systematically conducted as a link in the chain of steady reliability improvement activities. These advanced and/or continuous and steady technologies are one of most important factors for high reliability through the entire lifetime of a nuclear plant, including planning, design, construction, operation and maintenance stages. (author)

East-Asia nuclear/fossil power plant competitiveness

International Nuclear Information System (INIS)

Braun, Ch.

1996-01-01

The competitiveness of a new nuclear plant vs. a new oil or gas fired combined cycle plant or a coal fired plant in East-Asia, is reviewed in the paper. Both the nuclear and the fossil fired plants are evaluated as either utility financed or independent power producer (IPP) financed. Two types of advanced light water reactors (ALWRs) are considered in this paper, namely evolutionary ALWRs (1200 MWe size) and passive ALWRs (600 MWe class). A range of capital and total generation costs for each plant type is reported here. The comparison centers on three elements of overall competitiveness: generation costs, hard currency requirements, and employment requirements. Each of these aspects is considered perspective. Year-by-Year generation cost history over the plant lifetime is shown in some cases. It is found here that a utility financed evolutionary and passive ALWRs are broadly competitive with an IPP financed gas fired combined cycle plant and are more economic than oil fired combined cycle or a coal fired plant. A single unit evolutionary ALWR may have a 12 - 15 % capital cost advantage over a single passive ALWR then adjusted on a per KWe basis. Front-end hard currency requirements of a passive ALWR are 2.5 times higher than for a combined plant and evolutionary ALWRs requires 3.6 times higher up-front cost. However, on a lifetime basis, passive ALWR net hard currency requirements are two times lower than for a combined cycle plant. Evolutionary ALWR net hard currency requirements are three times over than those of a combined cycle plant. The effects of domestic vs. world price of fossil fuels on relative nuclear competitiveness are reviewed in this nuclear competitiveness paper. Employment requirements in an ALWR during both the construction period and lifetime operation, exceed the requirements for oil or gas fired plants by a factor of five. While contributing to overall plant cost, employment requirements can also be viewed as opportunity to increase national

Environmental impact of passive house nursery

Energy Technology Data Exchange (ETDEWEB)

Vares, S., Email: sirje.vares@vtt.fi

2012-06-15

It is often believed that reduction in energy use automatically leads to the total reduction of the carbon footprint and other emissions. To achieve better energy efficiency more raw materials may be needed not only for insulation and better windows but also for heating systems like ground source heat and solar panels. The use of advanced building systems increase the use of electricity and in winter where electricity production is already inadequate the additional stand-by power plants must be taken in use. These are not as effective as CHP plants for heat production. Moreover also the passive house structures can be produced in dozens ways and from many different materials which all have different service life, different need for maintenance and also different effect on the overall carbon footprint. Finally as the nursery is the overall concept, besides the building structures outdoor playgrounds and specific operations requiring day care trips, personnel commuting and waste treatment must be taken into account. (orig.)

Model-free adaptive control of advanced power plants

Science.gov (United States)

Cheng, George Shu-Xing; Mulkey, Steven L.; Wang, Qiang

2015-08-18

A novel 3-Input-3-Output (3.times.3) Model-Free Adaptive (MFA) controller with a set of artificial neural networks as part of the controller is introduced. A 3.times.3 MFA control system using the inventive 3.times.3 MFA controller is described to control key process variables including Power, Steam Throttle Pressure, and Steam Temperature of boiler-turbine-generator (BTG) units in conventional and advanced power plants. Those advanced power plants may comprise Once-Through Supercritical (OTSC) Boilers, Circulating Fluidized-Bed (CFB) Boilers, and Once-Through Supercritical Circulating Fluidized-Bed (OTSC CFB) Boilers.

Advanced control room design for nuclear power plants

International Nuclear Information System (INIS)

Scarola, K.

1987-01-01

The power industry has seen a continuous growth of size and complexity of nuclear power plants. Accompanying these changes have been extensive regulatory requirements resulting in significant construction, operation and maintenance costs. In response to related concerns raised by industry members, Combustion Engineering developed the NUPLEX 80 Advanced Control Room. The goal of NUPLEX 80 TM is to: reduce design and construction costs; increase plant safety and availability through improvements in the man-machine interface; and reduce maintenance costs. This paper provides an overview of the NUPLEX 80 Advanced Control Room and explains how the stated goals are achieved. (author)

Technical analysis of advanced wastewater-treatment systems for coal-gasification plants

Energy Technology Data Exchange (ETDEWEB)

1981-03-31

This analysis of advanced wastewater treatment systems for coal gasification plants highlights the three coal gasification demonstration plants proposed by the US Department of Energy: The Memphis Light, Gas and Water Division Industrial Fuel Gas Demonstration Plant, the Illinois Coal Gasification Group Pipeline Gas Demonstration Plant, and the CONOCO Pipeline Gas Demonstration Plant. Technical risks exist for coal gasification wastewater treatment systems, in general, and for the three DOE demonstration plants (as designed), in particular, because of key data gaps. The quantities and compositions of coal gasification wastewaters are not well known; the treatability of coal gasification wastewaters by various technologies has not been adequately studied; the dynamic interactions of sequential wastewater treatment processes and upstream wastewater sources has not been tested at demonstration scale. This report identifies key data gaps and recommends that demonstration-size and commercial-size plants be used for coal gasification wastewater treatment data base development. While certain advanced treatment technologies can benefit from additional bench-scale studies, bench-scale and pilot plant scale operations are not representative of commercial-size facility operation. It is recommended that coal gasification demonstration plants, and other commercial-size facilities that generate similar wastewaters, be used to test advanced wastewater treatment technologies during operation by using sidestreams or collected wastewater samples in addition to the plant's own primary treatment system. Advanced wastewater treatment processes are needed to degrade refractory organics and to concentrate and remove dissolved solids to allow for wastewater reuse. Further study of reverse osmosis, evaporation, electrodialysis, ozonation, activated carbon, and ultrafiltration should take place at bench-scale.

Modular construction approach for advanced nuclear plants

International Nuclear Information System (INIS)

Johnson, F.T.; Orr, R.S.; Boudreaux, C.P.

1988-01-01

Modular construction has been designated as one of the major features of the AP600 program, a small innovative 600-MW (electric) advanced light water reactor (ALWR) that is currently being developed by Westinghouse and its subcontractors. This program is sponsored by the US Department of Energy (DOE) in conjunction with several other DOE and Electric Power Research Institute ALWR programs. Two major objectives of the AP600 program are as follows: (1) to provide a cost of power competitive with other power generation alternatives; and (2) to provide a short construction schedule that can be met with a high degree of certainty. The AP600 plant addresses these objectives by providing a simplified plant design and an optimized plant arrangement that result in a significant reduction in the number and size of systems and components, minimizes the overall building volumes, and consequently reduces the required bulk quantities. However, only by adopting a modular construction approach for the AP600 can the full cost and schedule benefits be realized from the advances made in the plant systems design and plant arrangement. Modularization is instrumental in achieving both of the above objectives, but most of all, a total modularization approach is considered absolutely essential to ensure that an aggressive construction schedule can be met with a high degree of certainty

Recent advances in plant-herbivore interactions [version 1; referees: 2 approved

Directory of Open Access Journals (Sweden)

Deron E. Burkepile

2017-02-01

Full Text Available Plant-herbivore interactions shape community dynamics across marine, freshwater, and terrestrial habitats. From amphipods to elephants and from algae to trees, plant-herbivore relationships are the crucial link generating animal biomass (and human societies from mere sunlight. These interactions are, thus, pivotal to understanding the ecology and evolution of virtually any ecosystem. Here, we briefly highlight recent advances in four areas of plant-herbivore interactions: (1 plant defense theory, (2 herbivore diversity and ecosystem function, (3 predation risk aversion and herbivory, and (4 how a changing climate impacts plant-herbivore interactions. Recent advances in plant defense theory, for example, highlight how plant life history and defense traits affect and are affected by multiple drivers, including enemy pressure, resource availability, and the local plant neighborhood, resulting in trait-mediated feedback loops linking trophic interactions with ecosystem nutrient dynamics. Similarly, although the positive effect of consumer diversity on ecosystem function has long been recognized, recent advances using DNA barcoding to elucidate diet, and Global Positioning System/remote sensing to determine habitat selection and impact, have shown that herbivore communities are probably even more functionally diverse than currently realized. Moreover, although most diversity-function studies continue to emphasize plant diversity, herbivore diversity may have even stronger impacts on ecosystem multifunctionality. Recent studies also highlight the role of risk in plant-herbivore interactions, and risk-driven trophic cascades have emerged as landscape-scale patterns in a variety of ecosystems. Perhaps not surprisingly, many plant-herbivore interactions are currently being altered by climate change, which affects plant growth rates and resource allocation, expression of chemical defenses, plant phenology, and herbivore metabolism and behavior. Finally

Experimental investigation of a two-phase closed thermosyphon assembly for passive containment cooling system

Energy Technology Data Exchange (ETDEWEB)

Nam, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Sang Nyung [Kyunghee Univ., Gyeonggi-do (Korea, Republic of)

2017-06-15

After the Fukushima accident, increasing interest has been raised in passive safety systems that maintain the integrity of the containment building. To improve the reliability and safety of nuclear power plants, long-term passive cooling concepts have been developed for advanced reactors. In a previous study, the proposed design was based on an ordinary cylindrical Two-Phase Closed Thermosyphon (TPCT). The exact assembly size and number of TPCTs should be elaborated upon through accurate calculations based on experiments. While the ultimate goal is to propose an effective MPHP design for the PCCS and experimentally verify its performance, a TPCT assembly that was manufactured based on the conceptual design in this paper was tested.

Passive radon daughter dosimeters

International Nuclear Information System (INIS)

McElroy, R.G.C.; Johnson, J.R.

1986-03-01

On the basis of an extensive review of the recent literature concerning passive radon daughter dosimeters, we have reached the following conclusions: 1) Passive dosimeters for measuring radon are available and reliable. 2) There does not presently exist an acceptable passive dosimeter for radon daughters. There is little if any hope for the development of such a device in the foreseeable future. 3) We are pessimistic about the potential of 'semi-passive dosimeters' but are less firm about stating categorically that these devices cannot be developed into a useful radon daughter dosimeter. This report documents and justifies these conclusions. It does not address the question of the worker's acceptance of these devices because at the present time, no device is sufficiently advanced for this question to be meaningful. 118 refs

Single-tube condensation experiment in Passive Auxiliary Feedwater System of APR1400+

Energy Technology Data Exchange (ETDEWEB)

Shin, Chang Wook; No, Hee Cheon; Yun, Bong Yo; Jeon, Byong Guk [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2012-05-15

Conventional Korean nuclear power plants, Advanced Power Reactors (APR), are characterized by an active cooling system. However, Active cooling system may not prevent significant damage without any AC power source available for its operation as vividly illustrated through the recent Fukushima incident. In the APR1400+ to be designed, an independent passive cooling system was added in order to overcome the aforementioned shortcomings. In the Passive Auxiliary Feedwater System (PAFS), gravity force and density difference between steam and water are used. The system comprises of 240 condensation tubes to efficiently remove decay heat. Before applying the PAFS to APR1400+, the system's safety and heat removal performance must be verified. The present study experimentally evaluates the heat removal performance of a single tube in the PAFS. The objectives of SCOP (Single-tube Condensation experiment facility of PAFS) are the evaluation of the heat removal performance in the tube of the PAFS and database construction under various tube designs and test conditions. Reaching these objectives, we developed advanced measurement techniques for the amount of moisture, heat flux, and water film thickness.

Exposure of Lima bean leaves to volatiles from herbivore-induced conspecific plants results in emission of carnivore attractants: active or passive process?

NARCIS (Netherlands)

Choh, Y.; Shimoda, T.; Ozawa, R.; Dicke, M.; Takabayashi, J.

2004-01-01

There is increasing evidence that volatiles emitted by herbivore-damaged plants can cause responses in downwind undamaged neighboring plants, such as the attraction of carnivorous enemies of herbivores. One of the open questions is whether this involves an active (production of volatiles) or passive

Implementation of passive samplers for monitoring volatile organic compounds in ground water at the Kansas City Plant

International Nuclear Information System (INIS)

Gardner, F.G.; Korte, N.E.; Wilson-Nichols, M.J.; Baker, J.L.; Ramm, S.G.

1998-06-01

Passive sampling for monitoring volatile organic compounds (VOCs) has been suggested as a possible replacement to the traditional bailer method used at the Department of Energy Kansas City Plant (KCP) for routine groundwater monitoring. To compare methods, groundwater samples were collected from 19 KCP wells with VOC concentrations ranging from non-detectable to > 100,000 microg/L. Analysis of the data was conducted using means and medians of multiple measurements of TCE, 1,2-DCE, 1,1-DCE and VC. All 95% confidence intervals of these VOCs overlap, providing evidence that the two methods are similar. The study also suggests that elimination of purging and decontamination of sampling equipment reduces the labor required to sample by approximately 32%. Also, because the passive method generates no waste water, there are no associated disposal costs. The results suggest evidence to continue studies and efforts to replace traditional bailer methods with passive sampling at KCP based on cost and the similarity of the methods

Advanced configuration of hybrid passive filter for reactive power and harmonic compensation

OpenAIRE

Kececioglu, O. Fatih; Acikgoz, Hakan; Sekkeli, Mustafa

2016-01-01

Harmonics is one of the major power quality problems for power systems. The harmonics can be eliminated by power filters such as passive, active, and hybrid. In this study, a new passive filter configuration has been improved in addition to the existing passive filter configurations. Conventional hybrid passive filters are not successful to compensate rapidly changing reactive power demand. The proposed configure are capable of compensating both harmonics and reactive power at the same time. ...

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.

A Methodology for Modeling Nuclear Power Plant Passive Component Aging in Probabilistic Risk Assessment under the Impact of Operating Conditions, Surveillance and Maintenance Activities

Science.gov (United States)

Guler Yigitoglu, Askin

In the context of long operation of nuclear power plants (NPPs) (i.e., 60-80 years, and beyond), investigation of the aging of passive systems, structures and components (SSCs) is important to assess safety margins and to decide on reactor life extension as indicated within the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. In the traditional probabilistic risk assessment (PRA) methodology, evaluating the potential significance of aging of passive SSCs on plant risk is challenging. Although passive SSC failure rates can be added as initiating event frequencies or basic event failure rates in the traditional event-tree/fault-tree methodology, these failure rates are generally based on generic plant failure data which means that the true state of a specific plant is not reflected in a realistic manner on aging effects. Dynamic PRA methodologies have gained attention recently due to their capability to account for the plant state and thus address the difficulties in the traditional PRA modeling of aging effects of passive components using physics-based models (and also in the modeling of digital instrumentation and control systems). Physics-based models can capture the impact of complex aging processes (e.g., fatigue, stress corrosion cracking, flow-accelerated corrosion, etc.) on SSCs and can be utilized to estimate passive SSC failure rates using realistic NPP data from reactor simulation, as well as considering effects of surveillance and maintenance activities. The objectives of this dissertation are twofold: The development of a methodology for the incorporation of aging modeling of passive SSC into a reactor simulation environment to provide a framework for evaluation of their risk contribution in both the dynamic and traditional PRA; and the demonstration of the methodology through its application to pressurizer surge line pipe weld and steam generator tubes in commercial nuclear power plants. In the proposed methodology, a

In-plant measurements of gamma-ray transmissions for precise K-edge and passive assay of plutonium concentration and isotopic abundance in product solutions at the Tokai Reprocessing Plant

International Nuclear Information System (INIS)

Asakura, Y.; Kondo, I.; Masui, J.; Shoji, K.; Russo, P.A.; Hsue, S.T.; Sprinkle, J.K. Jr.; Johnson, S.S.

1982-01-01

A field test has been carried out for more than 2 years for determination of plutonium concentration by K-edge absorption densitometry and for determination of plutonium isotopic abundance by transmission-corrected passive gamma-ray spectrometry. This system was designed and built at Los Alamos National Laboratory and installed at the Tokai reprocessing plant of the Power Reactor and Nuclear Fuel Development Corporation as a part of the Tokai Advanced Safeguards Technology Exercise (TASTEX). For K-edge measurement of plutonium concentration, the transmissions at two discrete gamma-ray energies are measured using the 121.1- and 122.1-keV gamma rays from 75 Se and 57 Co. Intensities of the plutonium passive gamma rays in the energy regions between 38 and 51 keV and between 129 and 153 keV are used for determination of the isotopic abundances. More than 200 product solution samples have been measured in a timely fashion during these 2 years. The relative precisions and accuracies of the plutonium concentration measurement are shown to be within 0.6% (1 sigma) in these applications, and those for plutonium isotopic abundances are within 3% for 238 Pu, 0.4% for 239 Pu, 1.2% for 240 Pu, 1.3% for 241 Pu, and 7% for 242 Pu. The time required is 10 min for the concentration assay, 10 min for the isotopics assay, and about 15 min for handling procedures in the laboratory

Passive heat removal characteristics of SMART

Energy Technology Data Exchange (ETDEWEB)

Seo, Jae Kwang; Kang, Hyung Seok; Yoon, Joo Hyun; Kim, Hwan Yeol; Cho, Bong Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRS of the SMART has excellent passive heat removal characteristics. 2 refs., 4 figs., 1 tab. (Author)

Advanced man-machine system for nuclear power plants

International Nuclear Information System (INIS)

Masui, Takao; Naito, Norio; Kato, Kanji.

1990-01-01

Recent development of artificial intelligence(AI) seems to offer new possibility to strengthen the performance of the operator support system. From this point of view, a national project of Advanced Man-Machine System Development for Nuclear Power Plant (MMS-NPP) has been carried out since 1984 as 8-year project. This project aims at establishing advanced operator support functions which support operators in their knowledge-based behaviors and smoother interface with the system. This paper describes the role of MMS-NPP, the support functions and the main feature of the MMS-NPP detailed design with its focus placed on the realization methods using AI technology of the support functions for BWR and PWR plants. (author)

Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

Energy Technology Data Exchange (ETDEWEB)

Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

2015-06-19

This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

Advanced digital PWR plant protection system based on optimal estimation theory

International Nuclear Information System (INIS)

Tylee, J.L.

1981-04-01

An advanced plant protection system for the Loss-of-Fluid Test (LOFT) reactor plant is described and evaluated. The system, based on a Kalman filter estimator, is capable of providing on-line estimates of such critical variables as fuel and cladding temperature, departure from nucleate boiling ratio, and maximum linear heat generation rate. The Kalman filter equations are presented, as is a description of the LOFT plant dynamic model inherent in the filter. Simulation results demonstrate the performance of the advanced system

The link between exercise and titin passive stiffness.

Science.gov (United States)

Lalande, Sophie; Mueller, Patrick J; Chung, Charles S

2017-09-01

What is the topic of this review? This review focuses on how in vivo and molecular measurements of cardiac passive stiffness can predict exercise tolerance and how exercise training can reduce cardiac passive stiffness. What advances does it highlight? This review highlights advances in understanding the relationship between molecular (titin-based) and in vivo (left ventricular) passive stiffness, how passive stiffness modifies exercise tolerance, and how exercise training may be therapeutic for cardiac diseases with increased passive stiffness. Exercise can help alleviate the negative effects of cardiovascular disease and cardiovascular co-morbidities associated with sedentary behaviour; this may be especially true in diseases that are associated with increased left ventricular passive stiffness. In this review, we discuss the inverse relationship between exercise tolerance and cardiac passive stiffness. Passive stiffness is the physical property of cardiac muscle to produce a resistive force when stretched, which, in vivo, is measured using the left ventricular end diastolic pressure-volume relationship or is estimated using echocardiography. The giant elastic protein titin is the major contributor to passive stiffness at physiological muscle (sarcomere) lengths. Passive stiffness can be modified by altering titin isoform size or by post-translational modifications. In both human and animal models, increased left ventricular passive stiffness is associated with reduced exercise tolerance due to impaired diastolic filling, suggesting that increased passive stiffness predicts reduced exercise tolerance. At the same time, exercise training itself may induce both short- and long-term changes in titin-based passive stiffness, suggesting that exercise may be a treatment for diseases associated with increased passive stiffness. Direct modification of passive stiffness to improve exercise tolerance is a potential therapeutic approach. Titin passive stiffness itself may

Fireside corrosion of superheaters/reheaters in advanced power plants

Energy Technology Data Exchange (ETDEWEB)

Syed, A.U.; Simms, N.J.; Oakey, J.E. [Cranfield Univ. (United Kingdom). Energy Technology Centre

2010-07-01

The generation of increasing amounts of electricity while simultaneously reducing environmental emissions (CO{sub 2}, SO{sub 2}, NO{sub x} particles, etc) has become a goal for the power industry worldwide. Co-firing biomass and coal in new advanced pulverised fuel power plants is one route to address this issue, since biomass is regarded as a CO{sub 2} neutral fuel (i.e. CO{sub 2} uptake during its growth equals the CO{sub 2} emissions produced during its combustion) and such new advanced power plants operate at higher efficiencies than current plants as a result of using steam systems with high temperatures and pressures. However, co-firing has the potential to cause significant operational challenges for such power plants as amongst other issues, it will significantly change the chemistry of the deposits on the heat exchanger surfaces and the surrounding gas compositions. As a result these critical components can experience higher corrosion rates, and so shorter lives, causing increased operational costs, unless the most appropriate materials are selected for their construction. This paper reports the results of a series of 1000 hour laboratory corrosion tests that have been carried out in controlled atmosphere furnaces, to assess the effect of biomass/coal co-firing on the fireside corrosion of superheaters/reheaters. The materials used for the tests were one ferritic alloy (T92), two austenitic alloys (347HFG and HR3C) and one nickel based alloy (alloy 625). Temperatures of 600 and 650 C were used to represent the metal temperatures in advanced power plants. During these exposures, traditional mass change data were recorded as the samples were recoated with the simulated deposits. After these exposures, cross-sections through samples were prepared using standard metallographic techniques and then analysed using SEM/EDX. Pre-exposure micrometer and post-exposure image analyser measurements were used so that the metal wastage could be calculated. These data are

The passive safety systems of the Swr 1000

International Nuclear Information System (INIS)

Neumann, D.

2001-01-01

In recent years, a new boiling water reactor (BWR) plant called the SWR 1000 has been developed by Siemens on behalf of Germany's electric utilities. This new plant design concept incorporates the wide range of operating experience gained with German BWRs. The main objective behind developing the SWR 1000 was to design a plant with a rated electric output of approximately 1000 MW which would not only have a lower capital cost and lower power generating costs but would also provide a much higher level of nuclear safety compared to plants currently in operation. This safety-related goal has been met through, for example, the use of passive safety equipment. Passive systems make a significant contribution towards increasing the over-all level of plant safety due to the way in which they operate. They function solely accord-ing to basic laws of nature, such as gravity, and perform their designated functions with-out any need for electric power or other sources of external energy, or signals from instrumentation and control (I and C) equipment. The passive safety systems have been designed such that design basis accidents can be controlled using just these systems alone. However, the design concept of the SWR 1000 is nevertheless still based on the provision of active safety systems in addition to passive systems. (author)

Commercializing the next generation: the AP600 advanced simplified nuclear power plant

International Nuclear Information System (INIS)

Bruschi, H.J.

1994-01-01

Today, government and industry are working together on advanced nuclear power plant designs that take advantage of valuable lessons learned from the experience to date and promise to reconcile the demands of economic expansion with the laws of environmental protection. In the U.S., the Department of Energy (DOE) and the Electric Power Research Institute (EPRI) initiated a design certification program in 1989 to develop and commercialize advanced light water reactors (ALWRs) for the next round of power plant construction. Advanced, simplified technology is one approach under development to end the industry's search for a simpler, more forgiving, and less costly reactor. As part of this program, Westinghouse is developing the AP600, a new standard 600 MWe advanced, simplified plant. The design strikes a balance between the use of proven technology and new approaches. The result is a greatly streamlined plant that can meet safety regulations and reliability requirements, be economically competitive, and promote broader public confidence in nuclear energy. 1 fig

Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

International Nuclear Information System (INIS)

Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

1995-01-01

The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

Addressing the fundamental issues in reliability evaluation of passive safety of AP1000 for a comparison with active safety of PWR

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Yang Ming

2013-01-01

Passive safety systems adopted in advanced Pressurized Water Reactor (PWR), such as AP1000 and EPR, should attain higher reliability than the existing active safety systems of the conventional PWR. The objective of this study is to discuss the fundamental issues relating to the reliability evaluation of AP1000 passive safety systems for a comparison with the active safety systems of conventional PWR, based on several aspects. First, comparisons between conventional PWR and AP1000 are made from the both aspects of safety design and cost reduction. The main differences between these PWR plants exist in the configurations of safety systems: AP1000 employs the passive safety system while reducing the number of active systems. Second, the safety of AP1000 is discussed from the aspect of severe accident prevention in the event of large break loss of coolant accidents (LOCA). Third, detailed fundamental issues on reliability evaluation of AP1000 passive safety systems are discussed qualitatively by using single loop models of safety systems of both PWRs plants. Lastly, methodology to conduct quantitative estimation of dynamic reliability for AP1000 passive safety systems in LOCA condition is discussed, in order to evaluate the reliability of AP1000 in future by a success-path-based reliability analysis method (i.e., GO-FLOW). (author)

Planning of solar heated plant for low-energy houses and passive houses. An introduction; Planlegging av solvarmeanlegg for lavenergiboliger og passivhus. En introduksjon

Energy Technology Data Exchange (ETDEWEB)

Andresen, Inger

2008-07-01

This guide gives an introduction to the most important principles in planning and projecting of solar heated plant in low-energy houses and passive houses. It is written mainly for architects and consultants involved in housing projects with ambitions to achieve low-energy or passive house standard, but will also be of value for builders and others interested in the topic. (AG). 35 refs., 27 figs

Status of the IAEA coordinated research project on natural circulation phenomena, modelling, and reliability of passive systems that utilize natural circulation

International Nuclear Information System (INIS)

Reyes, J.N. Jr.; Cleveland, J.; Aksan, N.

2004-01-01

The International Atomic Energy Agency (IAEA) has established a Coordinated Research Project (CRP) titled ''Natural Circulation Phenomena, Modelling and Reliability of Passive Safety Systems that Utilize Natural Circulation. '' This work has been organized within the framework of the IAEA Department of Nuclear Energy's Technical Working Groups for Advanced Technologies for Light Water Reactors and Heavy Water Reactors (the TWG-LWR and the TWG-HWR). This CRP is part of IAEA's effort to foster international collaborations that strive to improve the economic performance of future water-cooled nuclear power plants while meeting stringent safety requirements. Thus far, IAEA has established 12 research agreements with organizations from industrialized Member States and 3 research contracts with organizations from developing Member States. The objective of the CRP is to enhance our understanding of natural circulation phenomena in water-cooled reactors and passive safety systems. The CRP participants are particularly interested in establishing a natural circulation and passive safety system thermal hydraulic database that can be used to benchmark computer codes for advanced reactor systems design and safety analysis. An important aspect of this CRP relates to developing methodologies to assess the reliability of passive safety systems in advanced reactor designs. This paper describes the motivation and objectives of the CRP, the research plan, and the role of each of the participating organizations. (author)

Passive containment system

International Nuclear Information System (INIS)

Kleimola, F.W.

1977-01-01

Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

Finite mixture models for sensitivity analysis of thermal hydraulic codes for passive safety systems analysis

Energy Technology Data Exchange (ETDEWEB)

Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)

2015-08-15

Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.

Licensed bases management for advanced nuclear plants

International Nuclear Information System (INIS)

O'Connell, J.; Rumble, E.; Rodwell, E.

2001-01-01

Prospective Advanced Nuclear Plant (ANP) owners must have high confidence that the integrity of the licensed bases (LB) of a plant will be effectively maintained over its life cycle. Currently, licensing engineers use text retrieval systems, database managers, and checklists to access, update, and maintain vast and disparate licensing information libraries. This paper describes the demonstration of a ''twin-engine'' approach that integrates a program from the emerging class of concept searching tools with a modern Product Data Management System (PDMS) to enhance the management of LB information for an example ANP design. (author)

Licensed bases management for advanced nuclear plants

Energy Technology Data Exchange (ETDEWEB)

O' Connell, J [Duke Engineering and Services, Marlborough, MA (United States); Rumble, E; Rodwell, E [EPRI, Palo Alto, CA (United States)

2001-07-01

Prospective Advanced Nuclear Plant (ANP) owners must have high confidence that the integrity of the licensed bases (LB) of a plant will be effectively maintained over its life cycle. Currently, licensing engineers use text retrieval systems, database managers, and checklists to access, update, and maintain vast and disparate licensing information libraries. This paper describes the demonstration of a ''twin-engine'' approach that integrates a program from the emerging class of concept searching tools with a modern Product Data Management System (PDMS) to enhance the management of LB information for an example ANP design. (author)

Advanced water chemistry management in power plants

International Nuclear Information System (INIS)

Regis, V.; Sigon, F.

1995-01-01

Advanced water management based on low external impact cycle chemistry technologies and processes, effective on-line water control and monitoring, has been verified to improve water utilization and to reduce plant liquid supply and discharge. Simulations have been performed to optimize system configurations and performances, with reference to a 4 x 320 MWe/once-through boiler/AVT/river cooled power plant, to assess the effectiveness of membrane separation technologies allowing waste water reuse, to enhance water management system design and to compare these solutions on a cost/benefit analysis. 6 refs., 3 figs., 3 tabs

Information management systems improve advanced plant design

International Nuclear Information System (INIS)

Turk, R.S.; Serafin, S.A.; Leckley, J.B.

1994-01-01

Computer-aided engineering tools are proving invaluable in both the design and operation of nuclear power plants. ABB Combustion Engineering's Advanced Light Water Reactor (ALWR) features a computerized Information Management System (IMS) as an integral part of the design. The System 80+IMS represents the most powerful information management tool for Nuclear Power Plants commercially available today. Developed by Duke Power Company specifically for use by nuclear power plant owner operators, the IMS consists of appropriate hardware and software to manage and control information flow for all plant related work or tasks in a systematic, consistent, coordinated and informative manner. A significant feature of this IMS is that it is primarily based on plant data. The principal design tool, PASCE (Plant Application and Systems from Combustion Engineering), is comprised of intelligent databases that describe the design and from which accurate plant drawings are created. Additionally the IMS includes, at its hub, a relational database management system and an associated document management system. The data-based approach and applications associated with the IMS were developed, and have proven highly effective, for plant modifications, configuration management, and operations and maintenance applications at Duke Power Company's operating nuclear plants. This paper presents its major features and benefits. 4 refs

Conventional and advanced exergetic analyses applied to a combined cycle power plant

International Nuclear Information System (INIS)

Petrakopoulou, Fontina; Tsatsaronis, George; Morosuk, Tatiana; Carassai, Anna

2012-01-01

Conventional exergy-based methods pinpoint components and processes with high irreversibilities. However, they lack certain insight. For a given advanced technological state, there is a minimum level of exergy destruction related to technological and/or economic constraints that is unavoidable. Furthermore, in any thermodynamic system, exergy destruction stems from both component interactions (exogenous) and component inefficiencies (endogenous). To overcome the limitations of the conventional analyses and to increase our knowledge about a plant, advanced exergy-based analyses have been developed. In this paper, a combined cycle power plant is analyzed using both conventional and advanced exergetic analyses. Except for the expander of the gas turbine system and the high-pressure steam turbine, most of the exergy destruction in the plant components is unavoidable. This unavoidable part is constrained by internal technological limitations, i.e. each component’s endogenous exergy destruction. High levels of endogenous exergy destruction show that component interactions do not contribute significantly to the thermodynamic inefficiencies. In addition, these inefficiencies are unavoidable to a large extent. With the advanced analysis, new improvement strategies are revealed that could not otherwise be found. -- Highlights: ► This is the first application of a complete advanced exergetic analysis to a complex power plant. ► In the three-pressure-level combined cycle power plant studied here, the improvement potential of the majority of the components is low, since most of the exergy destruction is unavoidable. ► Component interactions are generally of lower importance for the considered plant. ► Splitting the exogenous exergy destruction reveals one-to-one component interactions and improvement strategies. ► The advanced exergetic analysis is a necessary supplement to the conventional analysis in improving a complex system.

Passive heat removal in CANDU

International Nuclear Information System (INIS)

Hart, R.S.

1997-01-01

CANDU has a tradition of incorporating passive systems and passive components whenever they are shown to offer performance that is equal to or better than that of active systems, and to be economic. Examples include the two independent shutdown systems that employ gravity and stored energy respectively, the dousing subsystem of the CANDU 6 containment system, and the ability of the moderator to cool the fuel in the event that all coolant is lost from the fuel channels. CANDU 9 continues this tradition, incorporating a reserve water system (RWS) that increases the inventory of water in the reactor building and profiles a passive source of makeup water and/or heat sinks to various key process systems. The key component of the CANDU 9 reserve water system is a large (2500 cubic metres) water tank located at a high elevation in the reactor building. The reserve water system, while incorporating the recovery system functions, and the non-dousing functions of the dousing tank in CANDU 6, embraces other key systems to significantly extend the passive makeup/heat sink capability. The capabilities of the reserve water system include makeup to the steam generators secondary side if all other sources of water are lost; makeup to the heat transport system in the event of a leak in excess of the D 2 O makeup system capability; makeup to the moderator in the event of a moderator leak when the moderator heat sink is required; makeup to the emergency core cooling (ECC) system to assure NPSH to the ECC pumps during a loss of coolant accident (LOCA), and provision of a passive heat sink for the shield cooling system. Other passive designs are now being developed by AECL. These will be incorporated in future CANDU plants when their performance has been fully proven. This paper reviews the passive heat removal systems and features of current CANDU plants and the CANDU 9, and briefly reviews some of the passive heat removal concepts now being developed. (author)

Study on diverse passive decay heat removal approach

International Nuclear Information System (INIS)

Lin Qian; Si Shengyi

2012-01-01

One of the most important principles for nuclear safety is the decay heat removal in accidents. Passive decay heat removal systems are extremely helpful to enhance the safety. In currently design of many advanced nuclear reactors, kinds of passive systems are proposed or developed, such as the passive residual heat removal system, passive injection system, passive containment cooling system. These systems provide entire passive heat removal paths from core to ultimate heat sink. Various kinds of passive systems for decay heat removal are summarized; their common features or differences on heat removal paths and design principle are analyzed. It is found that, these passive decay heat removal paths are similarly common on and connected by several basic heat transfer modes and steps. By the combinations or connections of basic modes and steps, new passive decay heat removal approach or diverse system can be proposed. (authors)

Advanced Demonstration of Motion Correction for Ship-to-Ship Passive Inspections

Energy Technology Data Exchange (ETDEWEB)

Ziock, Klaus-Peter [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Boehnen, Chris Bensing [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ernst, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2013-09-30

Passive radiation detection is a key tool for detecting illicit nuclear materials. In maritime applications it is most effective against small vessels where attenuation is of less concern. Passive imaging provides: discrimination between localized (threat) and distributed (non-threat) sources, removal of background fluctuations due to nearby shorelines and structures, source localization to an individual craft in crowded waters, and background subtracted spectra. Unfortunately, imaging methods cannot be easily applied in ship-to-ship inspections because relative motion of the vessels blurs the results over many pixels, significantly reducing sensitivity. This is particularly true for the smaller water craft where passive inspections are most valuable. In this project we performed tests and improved the performance of an instrument (developed earlier under, “Motion Correction for Ship-to-Ship Passive Inspections”) that uses automated tracking of a target vessel in visible-light images to generate a 3D radiation map of the target vessel from data obtained using a gamma-ray imager.

Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

International Nuclear Information System (INIS)

Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

2011-01-01

Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

CATHENA Analysis Of Candu Advanced Passive Moderator Concept In Normal Operation Condition

International Nuclear Information System (INIS)

Alfa, Sudjatmi K

2001-01-01

In the CANDU - advanced passive moderator (APM) concept, the positive void reactivity is eliminated by reducing the density of the moderator. The simple model for the CANDU APM concept consists of the calandria, heat exchanger, pump, and a stabilizing tank, along with connecting piping. The calandria is divided into two parts, one part simulates the down area, while the other simulates up flow area. To demonstrate the thermalhydraulic behavior of the APM concept, Canadian algorithm for thermalhydraulic network analysis (CATHENA) code is used. The simulation for a pressure boundary condition of 300, 330 and 360 kPa and for water coolant mass flow rate boundary conditions of 2000 and 3000 kg/s respectively have been studied. Preliminary results show that there is boiling in the core, with vapor condensing in the heat exchanger. It is important to note, that the solution had not reached steady state when the boiling occurred

Advanced construction methods for new nuclear power plants

International Nuclear Information System (INIS)

Bilbao y Leon, Sama; Cleveland, John; Moon, Seong-Gyun; Tyobeka, Bismark

2009-01-01

The length of the construction and commissioning phases of nuclear power plants have historically been longer than for conventional fossil fuelled plants, often having a record of delays and cost overruns as a result from several factors including legal interventions and revisions of safety regulations. Recent nuclear construction projects however, have shown that long construction periods for nuclear power plants are no longer the norm. While there are several inter-related factors that influence the construction time, the use of advanced construction techniques has contributed significantly to reducing the construction length of recent nuclear projects. (author)

BWR plant advanced central control panel PODIA

International Nuclear Information System (INIS)

Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

1983-01-01

BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

Advanced nuclear plant control complex

International Nuclear Information System (INIS)

Scarola, K.; Jamison, S.; Manazir, R.M.; Rescorl, R.L.; Harmon, D.L.

1991-01-01

An advanced control room complex for a nuclear power plant, including a discrete indicator and alarm system which is nuclear qualified for rapid response to changes in plant parameters and a component control system which together provide a discrete monitoring and control capability at a panel in the control room. A separate data processing system, which need not be nuclear qualified, provides integrated and overview information to the control room and to each panel, through CRTs and a large, overhead integrated process status overview board. The discrete indicator and alarm system and the data processing system receive inputs from common plant sensors and validate the sensor outputs to arrive at a representative value of the parameter for use by the operator during both normal and accidental conditions, thereby avoiding the need for him to assimilate data from each sensor individually. The integrated process status board is at the apex of an information hierarchy that extends through four levels and provides access at each panel to the full display hierarchy. The control room panels are preferably of a modular construction, permitting the definition of inputs and outputs, the man machine interface, and the plant specific algorithms, to proceed in parallel with the fabrication of the panels, the installation of the equipment and the generic testing thereof. (author)

Plant Line Trial Evaluation of Viable Non-Chromium Passivation Systems for Electrolytin Tinplate, ETP (TRP 9911)

Energy Technology Data Exchange (ETDEWEB)

John A. Sinsel

2003-06-30

Plant trial evaluations have been completed for two zirconium-based, non-chromium passivation systems previously identified as possible alternatives to cathodic dichromate (CDC) passivation for electrolytic tinplate (ETP). These trials were done on a commercial electrolytic tin plating line at Weirton Steel and extensive evaluations of the materials resulting from these trials have been completed. All this was accomplished as a collaborative effort under the AISI Technology Roadmap Program and was executed by seven North American Tin Mill Products producers [Bethlehem Steel (now acquired by International Steel Group (ISG)), Dofasco Inc., National Steel (now acquired by U.S. Steel), U.S. Steel, USS-Posco, Weirton Steel, and Wheeling-Pittsburgh Steel] with funding partially from the Department of Energy (DOE) and partially on an equal cost sharing basis among project participants. The initial phases of this project involved optimization of application procedures for the non-chromium systems in the laboratories at Bethlehem Steel and Betz Dearborn followed by extensive testing with various lacquer formulations and food simulants in the laboratories at Valspar and PPG. Work was also completed at Dofasco and Weirton Steel to develop methods to prevent precipitation of insoluble solids as a function of time from the zirconate system. The results of this testing indicated that sulfide staining characteristics for the non-chromium passivation systems could be minimized but not totally eliminated and neither system was found to perform quite as good, in this respect, as the standard CDC system. As for the stability of zirconate treatment, a method was developed to stabilize this system for a sufficient period of time to conduct plant trial evaluations but, working with a major supplier of zirconium orthosulfate, a method for long term stabilization is still under development.

Advanced design nuclear power plants: Competitive, economical electricity. An analysis of the cost of electricity from coal, gas and nuclear power plants

International Nuclear Information System (INIS)

1992-06-01

This report presents an updated analysis of the projected cost of electricity from new baseload power plants beginning operation around the year 2000. Included in the study are: (1) advanced-design, standardized nuclear power plants; (2) low emissions coal-fired power plants; (3) gasified coal-fired power plants; and (4) natural gas-fired power plants. This analysis shows that electricity from advanced-design, standardized nuclear power plants will be economically competitive with all other baseload electric generating system alternatives. This does not mean that any one source of electric power is always preferable to another. Rather, what this analysis indicates is that, as utilities and others begin planning for future baseload power plants, advanced-design nuclear plants should be considered an economically viable option to be included in their detailed studies of alternatives. Even with aggressive and successful conservation, efficiency and demand-side management programs, some new baseload electric supply will be needed during the 1990s and into the future. The baseload generating plants required in the 1990s are currently being designed and constructed. For those required shortly after 2000, the planning and alternatives assessment process must start now. It takes up to ten years to plan, design, license and construct a new coal-fired or nuclear fueled baseload electric generating plant and about six years for a natural gas-fired plant. This study indicates that for 600-megawatt blocks of capacity, advanced-design nuclear plants could supply electricity at an average of 4.5 cents per kilowatt-hour versus 4.8 cents per kilowatt-hour for an advanced pulverized-coal plant, 5.0 cents per kilowatt-hour for a gasified-coal combined cycle plant, and 4.3 cents per kilowatt-hour for a gas-fired combined cycle combustion turbine plant

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

International Nuclear Information System (INIS)

Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-09-15

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

Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

Energy Technology Data Exchange (ETDEWEB)

Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

2013-05-17

This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and

Objectives for the development of advanced nuclear plants

International Nuclear Information System (INIS)

1993-01-01

The scope of this report is to reiterate the broad objectives for the development of advanced nuclear plants, to set forth some related subobjectives and to propose some universal goals for the development programmes. The majority of these can be categorized under the headings of enhancing safety, improving reliability and gaining better economics. These categories are used in the report followed by additional categories considered to be important within the global framework intended. Additional broad objectives appear unlikely but more subobjectives may become evident as time progresses and the need arises to express them in the intended more global framework. The goals also may change. The scope is therefore a set of objectives for development of advanced nuclear plants. The objectives are believed to be universally acceptable; they have been reviewed on that basis. 13 refs

Advanced Neutron Sources: Plant Design Requirements

International Nuclear Information System (INIS)

1990-07-01

The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW th , heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS

Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04

International Nuclear Information System (INIS)

2004-01-01

The 2004 International Congress on Advances in Nuclear Power Plants (ICAPP'04) provides a forum for the industry to exchange the latest ideas and research findings on nuclear plants from all perspectives. This conference builds on the success of last year's meeting held in Cordoba, Spain, and on the 2002 inaugural meeting held in Hollywood, Florida. Because of the hard work of many volunteers from around the world, ICAPP'04 has been successful in achieving its goal. More than 325 invited and contributed papers/presentations are part of this ICAPP. There are 5 invited plenary sessions and 70 technical sessions with contributed papers. The ICAPP'04 Proceedings contain almost 275 papers prepared by authors from 25 countries covering topics related to advances in nuclear power plant technology. The program by technical track deals with: 1 - Water-Cooled Reactor Programs and Issues (Status of All New Water-Cooled Reactor Programs; Advanced PWRs: Developmental Stage I; Advanced PWRs: Developmental Stage II; Advanced PWRs: Basic Design Stage; Advanced BWRs; Economics, Regulation, Licensing, and Construction; AP1000); 2 - High Temperature Gas Cooled Reactors (Pebble Bed Modular Reactors; Very High Temperature Reactors; HTR Fuels and Materials; Innovative HTRs and Fuel Cycles); 3 - Long Term Reactor Programs and Strategies (Supercritical Pressure Water Reactors; Lead-Alloy Fast Reactors; Sodium and Gas Fast Reactors; Status of Advanced Reactor Programs; Non-classical Reactor Concepts); 4 - Operation, Performance, and Reliability Management (Information Technology Effect on Plant Operation; Operation, Maintenance and Reliability; Improving Performance and Reducing O and M Costs; Plant Modernization and Retrofits); 5 - Plant Safety Assessment and Regulatory Issues (LOCA and non-LOCA Analysis Methodologies; LOCA and non-LOCA Plant Analyses; In-Vessel Retention; Containment Performance and Hydrogen Control; Advances in Severe Accident Analysis; Advances in Severe Accident

Advanced light water reactor plant

International Nuclear Information System (INIS)

Giedraityte, Zivile

2008-01-01

For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

Research of nuclear power plant in-service maintenance based on virtual reality

International Nuclear Information System (INIS)

Wang Yong; Kuang Weijun

2015-01-01

This paper presents a method of constructing nuclear power plant in-service maintenance virtual simulation scene and virtual maintenance process. Taking air baffles dismantling process of CAP1400(China Advanced Passive 1400) nuclear power plant as an instance, this paper discusses ergonomics, space analysis, time assessment based on virtual reality in the process of in-service maintenance. It demonstrates the advantage of using VR technology to design and verify in-service maintenance process of nuclear power plant compared to the conventional way. (author)

Advance in study of intelligent diagnostic method for nuclear power plant

International Nuclear Information System (INIS)

Zhou Gang; Yang Li

2008-01-01

The advance of research on the application of three types of intelligent diagnostic approach based on neural network (ANN), fuzzy logic and expert system to the operation status monitoring and fault diagnosis of nuclear power plant (NPP) was reviewed. The research status and characters on status monitoring and fault diagnosis approaches based on neural network, fuzzy logic and expert system for nuclear power plant were analyzed. The development trend of applied research on intelligent diagnostic approaches for nuclear power plant was explored. The analysis results show that the research achievements on intelligent diagnostic approaches based on fuzzy logic and expert system for nuclear power plant are not much relatively. The research of intelligent diagnostic approaches for nuclear power plant concentrate on the aspect of operation status monitoring and fault diagnosis based on neural networks for nuclear power plant. The advancing tendency of intelligent diagnostic approaches for nuclear power plant is the combination of various intelligent diagnostic approaches, the combination of neural network diagnostic approaches and other diagnostic approaches as well as multiple neural network diagnostic approaches. (authors)

AEA studies on passive decay heat removal in advanced reactors

International Nuclear Information System (INIS)

Lillington, J.N.

1994-01-01

The main objectives of the UK study were: to identify, describe and compare different types of systems proposed in current designs; to identify key scenarios in which passive decay heat removal systems play an important preventative or mitigative role; to assess the adequacy of the relevant experimental database; to assess the applicability and suitability of current generation models/codes for predicting passive decay heat removal; to assess the potential effectiveness of different systems in respect of certain key licensing questions

Licensed bases management for advanced nuclear plants

Energy Technology Data Exchange (ETDEWEB)

O' Connell, J. [Duke Engineering and Services, Marlborough, MA (United States); Rumble, E.; Rodwell, E. [EPRI, Palo Alto, CA (United States)

2001-07-01

Prospective Advanced Nuclear Plant (ANP) owners must have high confidence that the integrity of the licensed bases (LB) of a plant will be effectively maintained over its life cycle. Currently, licensing engineers use text retrieval systems, database managers, and checklists to access, update, and maintain vast and disparate licensing information libraries. This paper describes the demonstration of a ''twin-engine'' approach that integrates a program from the emerging class of concept searching tools with a modern Product Data Management System (PDMS) to enhance the management of LB information for an example ANP design. (author)

Passive systems for light water reactors

International Nuclear Information System (INIS)

Adinolfi, R.; Noviello, L.

1990-01-01

The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

Active or passive systems? The EPR approach

International Nuclear Information System (INIS)

Bonhomme, N.; Py, J.P.

1996-01-01

In attempting to review how EPR is contemplated to meet requirements applicable to future nuclear power plants, the authors indicate where they see the markets and the corresponding unit sizes for the EPR which is a generic key factor for competitiveness. There are no reason in industrialized countries, other than USA (where the investment and amortizing practices under control by Public Utility Commission are quite particular), not to build future plants in the 1000 to 1500 MWe range. Standardization, which has been actively applied all along the French program and for the Konvoi plants, does not prevent evolution and allows to concentrate large engineering effort in smooth realization of plants and achieve actual construction and commissioning without significant delays. In order to contribute to public trust renewal, a next generation of power reactors should be fundamentally less likely to incur serious accidents. To reach this goal the best of passive and active systems must be considered without forgetting that the most important source of knowledge is construction and operating experience. Criteria to assess passive systems investigated for possible implementation in the EPR, such as simplicity of design, impact on plant operation, safety and cost, are discussed. Examples of the principal passive systems investigated are described and reasons why they have been dropped after screening through the criteria are given. (author). 11 figs

Active or passive systems? The EPR approach

Energy Technology Data Exchange (ETDEWEB)

Bonhomme, N [Nuclear Power International, Cedex (France); Py, J P [FRAMATOME, Cedex (France)

1996-12-01

In attempting to review how EPR is contemplated to meet requirements applicable to future nuclear power plants, the authors indicate where they see the markets and the corresponding unit sizes for the EPR which is a generic key factor for competitiveness. There are no reason in industrialized countries, other than USA (where the investment and amortizing practices under control by Public Utility Commission are quite particular), not to build future plants in the 1000 to 1500 MWe range. Standardization, which has been actively applied all along the French program and for the Konvoi plants, does not prevent evolution and allows to concentrate large engineering effort in smooth realization of plants and achieve actual construction and commissioning without significant delays. In order to contribute to public trust renewal, a next generation of power reactors should be fundamentally less likely to incur serious accidents. To reach this goal the best of passive and active systems must be considered without forgetting that the most important source of knowledge is construction and operating experience. Criteria to assess passive systems investigated for possible implementation in the EPR, such as simplicity of design, impact on plant operation, safety and cost, are discussed. Examples of the principal passive systems investigated are described and reasons why they have been dropped after screening through the criteria are given. (author). 11 figs.

Second Law based definition of passivity/activity of devices

Science.gov (United States)

Sundqvist, Kyle M.; Ferry, David K.; Kish, Laszlo B.

2017-10-01

Recently, our efforts to clarify the old question, if a memristor is a passive or active device [1], triggered debates between engineers, who have had advanced definitions of passivity/activity of devices, and physicists with significantly different views about this seemingly simple question. This debate triggered our efforts to test the well-known engineering concepts about passivity/activity in a deeper way, challenging them by statistical physics. It is shown that the advanced engineering definition of passivity/activity of devices is self-contradictory when a thermodynamical system executing Johnson-Nyquist noise is present. A new, statistical physical, self-consistent definition based on the Second Law of Thermodynamics is introduced. It is also shown that, in a system with uniform temperature distribution, any rectifier circuitry that can rectify thermal noise must contain an active circuit element, according to both the engineering and statistical physical definitions.

High-sensitivity measurements for low-level TRU wastes using advanced passive neutron techniques

International Nuclear Information System (INIS)

Menlove, H.O.; Eccleston, G.W.

1992-01-01

In recent years, both passive- and active-neutron nondestructive assay (NDA) systems have been used to measure the uranium and plutonium content in 200-ell drums. Because of the heterogeneity of the wastes, representative sampling is not possible and NDA methods are preferred over destructive analysis. Active-neutron assay systems are used to measure the fissile isotopes such as 235 U, 23 Pu, and 241 Pu; the isotopic ratios are used to infer the total plutonium content and thus the specific disintegration rate. The active systems include 14-MeV-neutron (DT) generators with delayed-neutron counting, (D,T) generators with the differential die-away technique, and 252 Cf delayed-neutron shufflers. Passive assay systems (for example, segmented gamma-ray scanners)5 have used gamma-ray sessions, while others (for example, passive drum counters) used passive-neutron signals. We have developed a new passive-neutron measurement technique to improve the accuracy and sensitivity of the NDA of plutonium scrap and waste. This new 200-ell-drum assay system combines the classical NDA method of counting passive-neutron totals and coincidences from plutonium with the new features of ''add-a-source'' (AS) and multiplicity counting to improve the accuracy of matrix corrections and statistical techniques that improve the low-level detectability limits. This paper describes the improvements we have made in passive-neutron assay systems and compares the accuracies and detectability limits of passive- and active-neutron assay systems

Advanced digital computers, controls, and automation technologies for power plants: Proceedings

International Nuclear Information System (INIS)

Bhatt, S.C.

1992-08-01

This document is a compilation of the papers that were presented at an EPRI workshop on Advances in Computers, Controls, and Automation Technologies for Power Plants. The workshop, sponsored by EPRI's Nuclear Power Division, took place February 1992. It was attended by 157 representatives from electric utilities, equipment manufacturers, engineering consulting organizations, universities, national laboratories, government agencies and international utilities. More than 40% of the attendees were from utilities representing the majority group. There were 30% attendees from equipment manufacturers and the engineering consulting organizations. The participants from government agencies, universities, and national laboratories were about 10% each. The workshop included a keynote address, 35 technical papers, and vendor's equipment demonstrations. The technical papers described the state-of-the-art in the areas of recent utility digital upgrades such as digital feedwater controllers, steam generator level controllers, integrated plant computer systems, computer aided diagnostics, automated testing and surveillance and other applications. A group of technical papers presented the ongoing B ampersand W PWR integrated plant control system prototype developments with the triple redundant advanced digital control system. Several international papers from France, Japan and U.K. presented their programs on advance power plant design and applications. Significant advances in the control and automation technologies such as adaptive controls, self-tuning methods, neural networks and expert systems were presented by developers, universities, and national laboratories. Individual papers are indexed separately

Advanced robot vision system for nuclear power plants

International Nuclear Information System (INIS)

Onoguchi, Kazunori; Kawamura, Atsuro; Nakayama, Ryoichi.

1991-01-01

We have developed a robot vision system for advanced robots used in nuclear power plants, under a contract with the Agency of Industrial Science and Technology of the Ministry of International Trade and Industry. This work is part of the large-scale 'advanced robot technology' project. The robot vision system consists of self-location measurement, obstacle detection, and object recognition subsystems, which are activated by a total control subsystem. This paper presents details of these subsystems and the experimental results obtained. (author)

The Plant-Window System: A framework for an integrated computing environment at advanced nuclear power plants

International Nuclear Information System (INIS)

Wood, R.T.; Mullens, J.A.; Naser, J.A.

1997-01-01

Power plant data, and the information that can be derived from it, provide the link to the plant through which the operations, maintenance and engineering staff understand and manage plant performance. The extensive use of computer technology in advanced reactor designs provides the opportunity to greatly expand the capability to obtain, analyze, and present data about the plant to station personnel. However, to support highly efficient and increasingly safe operation of nuclear power plants, it is necessary to transform the vast quantity of available data into clear, concise, and coherent information that can be readily accessed and used throughout the plant. This need can be met by an integrated computer workstation environment that provides the necessary information and software applications, in a manner that can be easily understood and sued, to the proper users throughout the plan. As part of a Cooperative Research and Development Agreement with the Electric Power Research Institute, the Oak Ridge National laboratory has developed functional requirements for a Plant-Wide Integrated Environment Distributed On Workstations (Plant-Window) System. The Plant-Window System (PWS) can serve the needs of operations, engineering, and maintenance personnel at nuclear power stations by providing integrated data and software applications within a common computing environment. The PWS requirements identify functional capabilities and provide guidelines for standardized hardware, software, and display interfaces so as to define a flexible computing environment for both current generation nuclear power plants and advanced reactor designs

Feasibility of passive heat removal systems

Energy Technology Data Exchange (ETDEWEB)

Ashurko, Yu M [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1996-12-01

This paper presents a review of decay heat removal systems (DHRSs) used in liquid metal-cooled fast reactors (LMFRs). Advantages and the disadvantages of these DHRSs, extent of their passivity and prospects for their use in advanced fast reactor projects are analyzed. Methods of extending the limitations on the employment of individual systems, allowing enhancement in their effectiveness as safety systems and assuring their total passivity are described. (author). 10 refs, 10 figs.

Field observation of advance warning/advisory signage for passive railway crossings with restricted lateral sightline visibility: an experimental investigation.

Science.gov (United States)

Ward, N J; Wilde, G J

1995-04-01

This study evaluated a newly proposed series of signs intended for passive crossings with restrictions to lateral sightline visibility. These signs provide advance warning of a crossing and the restriction to lateral visibility. In addition, the signs advise motorists to come to a complete stop before crossing. Motorist behaviour was examined before and after installation of these signs at a rural passive crossing. A second site was observed in parallel to control partially for any confounding effects. Results indicated that motorists reduced speed and searched approach quadrants longer at points in the approachway after installation of the signs. However, there was no reliable increase in the number of motorists coming to complete stop, engaging in search behaviours, or classified as safe. The results are discussed in terms of reasons for the lack of compliance with the sign advisory.

Active and Passive Microrheology: Theory and Simulation

Science.gov (United States)

Zia, Roseanna N.

2018-01-01

Microrheological study of complex fluids traces its roots to the work of the botanist Robert Brown in the early nineteenth century. Indeed, passive microrheology and Brownian motion are one and the same. Once thought to reveal a fundamental life force, the phenomenon was ultimately leveraged by Einstein in proof of the atomic nature of matter ( Haw 2006 ). His work simultaneously paved the way for modern-day passive microrheology by connecting observable particle motion—diffusion—to solvent properties—the viscosity—via the well-known Stokes-Einstein relation. Advances in microscopy techniques in the last two decades have prompted extensions of the original model to generalized forms for passive probing of complex fluids. In the last decade, active microrheology has emerged as a means by which to interrogate the nonequilibrium behavior of complex fluids, in particular, the non-Newtonian rheology of dynamically heterogeneous and microscopically small systems. Here we review theoretical and computational approaches and advances in both passive and active microrheology, with a focus on the extent to which these techniques preserve the connection between single-particle motion and flow properties, as well as the rather surprising recovery of non-Newtonian flow behavior observed in bulk rheology.

Advanced smart tungsten alloys for a future fusion power plant

Science.gov (United States)

Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch; Rasinski, M.; Kreter, A.; Tan, X.; Schmitz, J.; Mao, Y.; Coenen, J. W.; Bram, M.; Gonzalez-Julian, J.

2017-06-01

The severe particle, radiation and neutron environment in a future fusion power plant requires the development of advanced plasma-facing materials. At the same time, the highest level of safety needs to be ensured. The so-called loss-of-coolant accident combined with air ingress in the vacuum vessel represents a severe safety challenge. In the absence of a coolant the temperature of the tungsten first wall may reach 1200 °C. At such a temperature, the neutron-activated radioactive tungsten forms volatile oxide which can be mobilized into atmosphere. Smart tungsten alloys are being developed to address this safety issue. Smart alloys should combine an acceptable plasma performance with the suppressed oxidation during an accident. New thin film tungsten-chromium-yttrium smart alloys feature an impressive 105 fold suppression of oxidation compared to that of pure tungsten at temperatures of up to 1000 °C. Oxidation behavior at temperatures up to 1200 °C, and reactivity of alloys in humid atmosphere along with a manufacturing of reactor-relevant bulk samples, impose an additional challenge in smart alloy development. First exposures of smart alloys in steady-state deuterium plasma were made. Smart tungsten-chroimium-titanium alloys demonstrated a sputtering resistance which is similar to that of pure tungsten. Expected preferential sputtering of alloying elements by plasma ions was confirmed experimentally. The subsequent isothermal oxidation of exposed samples did not reveal any influence of plasma exposure on the passivation of alloys.

Testing of a passive autocatalytic recombiner in the Surtsey facility

International Nuclear Information System (INIS)

Blanchat, T.K.; Malliakos, A.C.

2000-01-01

Passive autocatalytic recombiners (PARs) have been under consideration in the US as a combustible gas control system in operating plants and advanced light water reactor containments for design-basis accidents. Here, performance tests of a scaled passive autocatalytic recombiner (PAR) were performed in the Surtsey test vessel at Sandia National laboratories. Measured hydrogen depletion rate data were obtained and compared with previous work. Depletion rate is most likely proportional to PAR scale. PAR performance in steamy environments (with and without hydrophobic coating) was investigated. The tests determined that the PAR startup delay times decrease with increasing hydrogen concentrations in steamy environments. Tests with placement of the PAR near a wall (as opposed to a center location) yielded reduced depletion rates. Tests at low oxygen concentrations also showed a reduced recombination rate. The PAR repeatedly ignited hydrogen at ∼6 mol% concentration with a catalyst temperature near 940 K. Velocity data at the PAR exhaust were used to calculate the volumetric flow rate through the PAR as a function of the vessel hydrogen concentration

French concepts of ''passive safety''

International Nuclear Information System (INIS)

Dennielou, Y.; Serret, M.

1990-01-01

N 4 model, the French 1400 MW PWR of the 90's, exhibits many advanced features. As far as safety is concerned, the fully computerized control room design takes advantage of the operating experience feedback and largely improves the man machine interface. New post-accident procedures have been developed (the so-called ''physical states oriented procedures''). A complete consistent set of ''Fundamental Safety Rules'' have been issued. This however doesn't imply any significant modification of standard PWR with regard to the passive aspects of safety systems or functions. Nevertheless, traditional PWR safety systems largely use passive aspects: natural circulation, reactivity coefficients, gravity driven control rods, injection accumulators, so on. Moreover, probability calculations allow for comparison between the respective contributions of passive and of active failures. In the near future, eventual options of future French PWRs to be commissioned after 2000 will be evaluated; simplification, passive and forgiving aspects of safety systems will be thoroughly considered. (author)

Future generations of CANDU: advantages and development with passive safety

International Nuclear Information System (INIS)

Duffey, R. B.

2006-01-01

Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

Design and development of self-powered sensors on wireless sensor network for standalone plant critical data management during SBO and beyond design basis events

International Nuclear Information System (INIS)

Aparna, J.; Dulera, I.V.; Rama Rao, A.; Vijayan, P.K.

2015-01-01

Advanced reactors are designed with an aim of maximum safety, optimized fuel utilization and effective system design. Safety aspects in reactor designs are being viewed for all possible vulnerabilities, and as a result, robust self-regulating passive safety features have been favored in Gen IV and advanced reactor designs. In addition to passive systems, the accidents scenarios at Fukushima indicate the dire need of reliable and stand-alone self-powered sensors, for monitoring plant critical parameters for effective damage control actions. There is a strong need for plant critical data management and situation awareness during the unavailability of all conventional power sources in a nuclear power plant, during extended station blackout (SBO) conditions. These self-powered sensors would assist the operators in managing events like SBO and help in containing any Beyond Design Basis Events (BDBE) conditions, well away from the public domain

Advanced Passivation Technology and Loss Factor Minimization for High Efficiency Solar Cells.

Science.gov (United States)

Park, Cheolmin; Balaji, Nagarajan; Jung, Sungwook; Choi, Jaewoo; Ju, Minkyu; Lee, Seunghwan; Kim, Jungmo; Bong, Sungjae; Chung, Sungyoun; Lee, Youn-Jung; Yi, Junsin

2015-10-01

High-efficiency Si solar cells have attracted great attention from researchers, scientists, photovoltaic (PV) industry engineers for the past few decades. With thin wafers, surface passivation becomes necessary to increase the solar cells efficiency by overcoming several induced effects due to associated crystal defects and impurities of c-Si. This paper discusses suitable passivation schemes and optimization techniques to achieve high efficiency at low cost. SiNx film was optimized with higher transmittance and reduced recombination for using as an effective antireflection and passivation layer to attain higher solar cell efficiencies. The higher band gap increased the transmittance with reduced defect states that persisted at 1.68 and 1.80 eV in SiNx films. The thermal stability of SiN (Si-rich)/SiN (N-rich) stacks was also studied. Si-rich SiN with a refractive index of 2.7 was used as a passivation layer and N-rich SiN with a refractive index of 2.1 was used for thermal stability. An implied Voc of 720 mV with a stable lifetime of 1.5 ms was obtained for the stack layer after firing. Si-N and Si-H bonding concentration was analyzed by FTIR for the correlation of thermally stable passivation mechanism. The passivation property of spin coated Al2O3 films was also investigated. An effective surface recombination velocity of 55 cm/s with a high density of negative fixed charges (Qf) on the order of 9 x 10(11) cm(-2) was detected in Al2O3 films.

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

International Nuclear Information System (INIS)

Monteleone, S.

1994-04-01

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. 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, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors

Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

Energy Technology Data Exchange (ETDEWEB)

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

1994-04-01

This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. 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, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

Advanced fission and fossil plant economics-implications for fusion

International Nuclear Information System (INIS)

Delene, J.G.

1994-01-01

In order for fusion energy to be a viable option for electric power generation, it must either directly compete with future alternatives or serve as a reasonable backup if the alternatives become unacceptable. This paper discusses projected costs for the most likely competitors with fusion power for baseload electric capacity and what these costs imply for fusion economics. The competitors examined include advanced nuclear fission and advanced fossil-fired plants. The projected costs and their basis are discussed. The estimates for these technologies are compared with cost estimates for magnetic and inertial confinement fusion plants. The conclusion of the analysis is that fusion faces formidable economic competition. Although the cost level for fusion appears greater than that for fission or fossil, the costs are not so high as to preclude fusion's potential competitiveness

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)

The advanced main control console for next japanese PWR plants

International Nuclear Information System (INIS)

Tsuchiya, A.; Ito, K.; Yokoyama, M.

2001-01-01

The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

Carbon steel protection in G.S. (Girlder sulfide) heavy water fabrication plants. Control of iron content at the final stage of passivation. Pt. 10

International Nuclear Information System (INIS)

Rojo, E.A.

1991-01-01

This paper is part of a series which corresponds to the carbon steel behaviour as construction material for Girlder sulfide (G.S.) heavy water plants. The present work analyses the iron concentration study during passivation in the passivating fluid. At the beginning, during the formation of the most soluble sulfide -that is the mackinawite-, the iron concentration reaches more than 10 ppm. After some days, this iron concentration begins to decrease up to its stabilization under 0.1 ppm. This process, which occurs in the 9th. and 11th days, indicates that passivation is over, and that a pyrite and pyrrhotite-pyrite layer exists on the iron. Some differences exist between the results obtained and those previsible for the iron sulfides solubilities. In spite of these difficulties, the procedure is perfectly adequate to judge the passivation final stage. (Author) [es

An Assessment of Wind Plant Complex Flows Using Advanced Doppler Radar Measurements

Science.gov (United States)

Gunter, W. S.; Schroeder, J.; Hirth, B.; Duncan, J.; Guynes, J.

2015-12-01

As installed wind energy capacity continues to steadily increase, the need for comprehensive measurements of wind plant complex flows to further reduce the cost of wind energy has been well advertised by the industry as a whole. Such measurements serve diverse perspectives including resource assessment, turbine inflow and power curve validation, wake and wind plant layout model verification, operations and maintenance, and the development of future advanced wind plant control schemes. While various measurement devices have been matured for wind energy applications (e.g. meteorological towers, LIDAR, SODAR), this presentation will focus on the use of advanced Doppler radar systems to observe the complex wind flows within and surrounding wind plants. Advanced Doppler radars can provide the combined advantage of a large analysis footprint (tens of square kilometers) with rapid data analysis updates (a few seconds to one minute) using both single- and dual-Doppler data collection methods. This presentation demonstrates the utility of measurements collected by the Texas Tech University Ka-band (TTUKa) radars to identify complex wind flows occurring within and nearby operational wind plants, and provide reliable forecasts of wind speeds and directions at given locations (i.e. turbine or instrumented tower sites) 45+ seconds in advance. Radar-derived wind maps reveal commonly observed features such as turbine wakes and turbine-to-turbine interaction, high momentum wind speed channels between turbine wakes, turbine array edge effects, transient boundary layer flow structures (such as wind streaks, frontal boundaries, etc.), and the impact of local terrain. Operational turbine or instrumented tower data are merged with the radar analysis to link the observed complex flow features to turbine and wind plant performance.

Passive containment system in high earthquake motion

International Nuclear Information System (INIS)

Kleimola, F.W.; Falls, O.B. Jr.

1977-01-01

High earthquake motion necessitates major design modifications in the complex of plant structures, systems and components in a nuclear power plant. Distinctive features imposed by seismic category, safety class and quality classification requirements for the high seismic ground acceleration loadings significantly reflect in plant costs. The design features in the Passive Containment System (PCS) responding to high earthquake ground motion are described

Passive restoration augments active restoration in deforested landscapes: the role of root suckering adjacent to planted stands of Acacia koa

Science.gov (United States)

Paul G. Scowcroft; Justin T. Yeh

2013-01-01

Active forest restoration in Hawaii’s Hakalau Forest National Wildlife Refuge has produced a network of Acacia koa tree corridors and islands in deforested grasslands. Passive restoration by root suckering has potential to expand tree cover and close gaps between planted stands. This study documents rates of encroachment into grassland, clonal...

Climate-associated phenological advances in bee pollinators and bee-pollinated plants

Science.gov (United States)

Bartomeus, Ignasi; Ascher, John S.; Wagner, David; Danforth, Bryan N.; Colla, Sheila; Kornbluth, Sarah; Winfree, Rachael

2011-01-01

The phenology of many ecological processes is modulated by temperature, making them potentially sensitive to climate change. Mutualistic interactions may be especially vulnerable because of the potential for phenological mismatching if the species involved do not respond similarly to changes in temperature. Here we present an analysis of climate-associated shifts in the phenology of wild bees, the most important pollinators worldwide, and compare these shifts to published studies of bee-pollinated plants over the same time period. We report that over the past 130 y, the phenology of 10 bee species from northeastern North America has advanced by a mean of 10.4 ± 1.3 d. Most of this advance has taken place since 1970, paralleling global temperature increases. When the best available data are used to estimate analogous rates of advance for plants, these rates are not distinguishable from those of bees, suggesting that bee emergence is keeping pace with shifts in host-plant flowering, at least among the generalist species that we investigated. PMID:22143794

National nuclear power planning of China and advanced reactor

International Nuclear Information System (INIS)

Qian Jihui

1990-01-01

The necessity of investigation on the trends of advanced reactor technology all over the world is elabrated while China is going to set up its long-term national nuclear power programme. In author's opinion, thermal reactor power plants will have a quite long period development in the next century and a new trend of second generation NPPs might emerge in the beginning of next century. These new generation advanced reactors are characterized with new design concepts based on the inherent or passive safety features. Among them, most promising ones are those of AP-600 and MHTGR. Chinese experts are paying special attention to and closely following these two directions

Passive safety systems for decay heat removal of MRX

Energy Technology Data Exchange (ETDEWEB)

Ochiai, M; Iida, H; Hoshi, T [Japan Atomic Energy Research Inst., Ibaraki (Japan). Nuclear Ship System Lab.

1996-12-01

The MRX (marine Reactor X) is an advanced marine reactor, its design has been studied in Japan Atomic Energy Research Institute. It is characterized by four features, integral type PWR, in-vessel type control rod drive mechanisms, water-filled containment vessel and passive decay heat removal system. A water-filled containment vessel is of great advantage since it ensures compactness of a reactor plant by realizing compact radiation shielding. The containment vessel also yields passive safety of MRX in the event of a LOCA by passively maintaining core flooding without any emergency water injection. Natural circulation of water in the vessels (reactor and containment vessels) is one of key factors of passive decay heat removal systems of MRX, since decay heat is transferred from fuel rods to atmosphere by natural circulation of the primary water, water in the containment vessel and thermal medium in heat pipe system for the containment vessel water cooling in case of long terms cooling after a LOCA as well as after reactor scram. Thus, the ideal of water-filled containment vessel is considered to be very profitable and significant in safety and economical point of view. This idea is, however, not so familiar for a conventional nuclear system, so experimental and analytical efforts are carried out for evaluation of hydrothermal behaviours in the reactor pressure vessel and in the containment vessel in the event of a LOCA. The results show the effectiveness of the new design concept. Additional work will also be conducted to investigate the practical maintenance of instruments in the containment vessel. (author). 4 refs, 9 figs, 2 tabs.

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)

Waste-to-energy advanced cycles and new design concepts for efficient power plants

CERN Document Server

Branchini, Lisa

2015-01-01

This book provides an overview of state-of-the-art technologies for energy conversion from waste, as well as a much-needed guide to new and advanced strategies to increase Waste-to-Energy (WTE) plant efficiency. Beginning with an overview of municipal solid waste production and disposal, basic concepts related to Waste-To-Energy conversion processes are described, highlighting the most relevant aspects impacting the thermodynamic efficiency of WTE power plants. The pervasive influences of main steam cycle parameters and plant configurations on WTE efficiency are detailed and quantified. Advanc

Advanced I and C systems for nuclear power plants

International Nuclear Information System (INIS)

Bock, H.W.; Graf, A.; Hofmann, H.

1995-01-01

Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally, specific requirements coming from the nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. The main features of these systems are the clear task related architecture with adaptable redundancy, the consequent application of standards for interfaces and communication, comprehensive tools for easy design and service and a highly ergonomic screen based man-machine-interface. The engineering tasks are supported by an integrated engineering system, which has the capacity for design, test and diagnosis of all I and C functions and the related equipment. TELEPERM XP is designed to optimally perform all automatic functions, which require no nuclear specific qualification. This includes all sequences and closed-loop controls as well as most man-machine-interface functions. TELEPERM XS is designed for all control tasks which require a nuclear specific qualification. This especially includes all functions to initiate automatic countermeasures to prevent or to cope with accidents. By use of the complementary I and C systems TELEPERM XP and TELEPERM XS, economical as well as advanced plant automation and man-machine-interfaces can be implemented into Nuclear Plants, assuring the compliance with the relevant international safety standards. (author). 10 figs

Advanced I and C systems for nuclear power plants

International Nuclear Information System (INIS)

Bock, H.W.; Graf, A.; Hofmann, H.

1997-01-01

Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. The main features of these systems are a clear task related architecture with adaptable redundancy, a consequent application of standards for interfaces and communication, comprehensive tools for easy design and service and a highly ergonomic screen based man-machine-interface. The engineering tasks are supported by an integrated engineering system, which has the capacity for design, test and diagnosis of all I and C functions and the related equipment. TELEPERM XP is designed to optimally perform all automatic functions, which require no nuclear specific qualification. This includes all sequences and closed-loop controls as well as most man-machine-interface functions. TELEPERM XS is designed for all control tasks which require a nuclear specific qualification. This especially includes all function to initiated automatic countermeasures to prevent or to cope with accidents. By use of the complementary I and C systems TELEPERM XP and TELEPERM XS, advanced and likewise economical plant automation and man-machine-interfaces can be implemented into Nuclear Power Plant, assuring compliance with the relevant international safety standards. (author). 10 figs

Advanced I and C systems for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Bock, H W; Graf, A; Hofmann, H [Siemens AG, Erlangen (Germany)

1997-07-01

Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. The main features of these systems are a clear task related architecture with adaptable redundancy, a consequent application of standards for interfaces and communication, comprehensive tools for easy design and service and a highly ergonomic screen based man-machine-interface. The engineering tasks are supported by an integrated engineering system, which has the capacity for design, test and diagnosis of all I and C functions and the related equipment. TELEPERM XP is designed to optimally perform all automatic functions, which require no nuclear specific qualification. This includes all sequences and closed-loop controls as well as most man-machine-interface functions. TELEPERM XS is designed for all control tasks which require a nuclear specific qualification. This especially includes all function to initiated automatic countermeasures to prevent or to cope with accidents. By use of the complementary I and C systems TELEPERM XP and TELEPERM XS, advanced and likewise economical plant automation and man-machine-interfaces can be implemented into Nuclear Power Plant, assuring compliance with the relevant international safety standards. (author). 10 figs.

Proceedings of the 18th national passive solar conference. Volume 18

International Nuclear Information System (INIS)

Burley, S.; Arden, M.E.

1993-01-01

The American Solar Energy Society conducts the National Solar Energy Conference as an annual forum for exchange of information about advances in solar energy technologies, programs, and concepts. The SOLAR 93 conference presented papers on the following topics: passive design tools; passive performance; building case studies; passive components, construction and glazing; daylighting; passive cooling; sustainability theory; sustainability projects; vernacular architecture; emerging architecture; and education. A total of forty-nine papers were indexed separately for the data base

Study on diverse passive decay heat removal approach and principle

International Nuclear Information System (INIS)

Lin Qian; Si Shengyi

2012-01-01

Decay heat removal in post-accident is one of the most important aspects concerned in the reactor safety analysis. Passive decay heat removal approach is used to enhance nuclear safety. In advanced reactors, decay heat is removed by multiple passive heat removal paths through core to ultimate heat sink by passive residual heat removal system, passive injection system, passive containment cooling system and so on. Various passive decay heat removal approaches are summarized in this paper, the common features and differences of their heat removal paths are analyzed, and the design principle of passive systems for decay heat removal is discussed. It is found that. these decay heat removal paths is combined by some basic heat transfer processes, by the combination of these basic processes, diverse passive decay heat removal approach or system design scheme can be drawn. (authors)

Confirmatory analysis of the AP1000 passive residual heat removal heat exchanger with 3-D computational fluid dynamic analysis

International Nuclear Information System (INIS)

Schwall, James R.; Karim, Naeem U.; Thakkar, Jivan G.; Taylor, Creed; Schulz, Terry; Wright, Richard F.

2006-01-01

The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power up-rate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model

Improvement of environmental aspects of thermal power plant operation by advanced control concepts

Directory of Open Access Journals (Sweden)

Mikulandrić Robert

2012-01-01

Full Text Available The necessity of the reduction of greenhouse gas emissions, as formulated in the Kyoto Protocol, imposes the need for improving environmental aspects of existing thermal power plants operation. Improvements can be reached either by efficiency increment or by implementation of emission reduction measures. Investments in refurbishment of existing plant components or in plant upgrading by flue gas desulphurization, by primary and secondary measures of nitrogen oxides reduction, or by biomass co-firing, are usually accompanied by modernisation of thermal power plant instrumentation and control system including sensors, equipment diagnostics and advanced controls. Impact of advanced control solutions implementation depends on technical characteristics and status of existing instrumentation and control systems as well as on design characteristics and actual conditions of installed plant components. Evaluation of adequacy of implementation of advanced control concepts is especially important in Western Balkan region where thermal power plants portfolio is rather diversified in terms of size, type and commissioning year and where generally poor maintenance and lack of investments in power generation sector resulted in high greenhouse gases emissions and low efficiency of plants in operation. This paper is intended to present possibilities of implementation of advanced control concepts, and particularly those based on artificial intelligence, in selected thermal power plants in order to increase plant efficiency and to lower pollutants emissions and to comply with environmental quality standards prescribed in large combustion plant directive. [Acknowledgements. This paper has been created within WBalkICT - Supporting Common RTD actions in WBCs for developing Low Cost and Low Risk ICT based solutions for TPPs Energy Efficiency increasing, SEE-ERA.NET plus project in cooperation among partners from IPA SA - Romania, University of Zagreb - Croatia and Vinca

Electropotential measurements of passivation and corrosion of steel coupons

International Nuclear Information System (INIS)

Petit, G.S.; Wright, R.R.

1977-02-01

There is considerable interest at the Oak Ridge Gaseous Diffusion Plant (ORGDP) in the preparation of mild steel to resist corrosion (passivation) both in moist air and uranium hexafluoride (UF 6 ) environments. Steel prepared by the usual procedures to prevent rusting, such as oiling, plastic coating, painting, or phosphating, cannot be used in the presence of UF 6 . Tests have shown that a chromate treatment or an ammoniacal citrate treatment for passivation are effective. The electropotential behavior of these two passivation treatments is described. The initial electropotential measurement, when compared to that of an unpassivated coupon, gives the electropotential degree in volts of passivation. Continual exposure in the test, when compared to the unpassivated coupon, gives a profile of the durability of the passivation film. The chromate passivation treatment was slightly superior to the citrate passivation

Development of advanced secondary chemistry monitoring system for Korea nuclear power plants

International Nuclear Information System (INIS)

Lee, Sang Hak; Kim, Chung Tae

1997-01-01

Water chemistry control is one of the most important tasks in order to maintain the reliability of plant equipments and extend the operating life of the plant. KEPCO and KOPEC developed a computerized tool for this purpose -ASCMS (advanced secondary chemistry monitoring system) which is able to monitor and diagnose the secondary water chemistry. A prototype system had been installed at KORI 3 nuclear power plant since April 1993 in order to evaluate the system performance. After the implementation of enhancements identified during the testing of the prototype, we have developed the advanced secondary monitoring system, ASCMS which is installed at 5 nuclear power plants and has been under operations since April 1997. The ASCMS comprises PC subsystem designed for data acquisition, data analysis, and data diagnosis. The ASCMS will provide overall information related to steam generator secondary side water chemistry problems and improve plant availability, reduce radiation exposure to workers, and reduce operating and maintenance costs. 6 figs

Passive scalar transport in peripheral regions of random flows

International Nuclear Information System (INIS)

Chernykh, A.; Lebedev, V.

2011-01-01

We investigate statistical properties of the passive scalar mixing in random (turbulent) flows assuming its diffusion to be weak. Then at advanced stages of the passive scalar decay, its unmixed residue is primarily concentrated in a narrow diffusive layer near the wall and its transport to the bulk goes through the peripheral region (laminar sublayer of the flow). We conducted Lagrangian numerical simulations of the process for different space dimensions d and revealed structures responsible for the transport, which are passive scalar tongues pulled from the diffusive boundary layer to the bulk. We investigated statistical properties of the passive scalar and of the passive scalar integrated along the wall. Moments of both objects demonstrate scaling behavior outside the diffusive boundary layer. We propose an analytic scheme for the passive scalar statistics, explaining the features observed numerically.

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

1989-10-01

A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

Studies on the behaviour of a passive containment cooling system for the Indian advanced heavy water reactor

International Nuclear Information System (INIS)

Maheshwari, N.K.; Saha, D.; Chandraker, D.K.; Kakodkar, A.; Venkat Raj, V.

2001-01-01

A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser's working is the potential degradation of immersed condenser's performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to understand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V 1 and V 2 reached almost steady value. With varying energy input rate V 1 pressure dropped below the pressure in V 2 . The system could efficiently purge air from V 1 to V 2 . The paper deals with the details of the tests conducted and the results obtained. (orig.) [de

Recent Molecular Advances on Downstream Plant Responses to Abiotic Stress

Directory of Open Access Journals (Sweden)

Cláudia Regina Batista de Souza

2012-07-01

Full Text Available Abiotic stresses such as extremes of temperature and pH, high salinity and drought, comprise some of the major factors causing extensive losses to crop production worldwide. Understanding how plants respond and adapt at cellular and molecular levels to continuous environmental changes is a pre-requisite for the generation of resistant or tolerant plants to abiotic stresses. In this review we aimed to present the recent advances on mechanisms of downstream plant responses to abiotic stresses and the use of stress-related genes in the development of genetically engineered crops.

Passive cooling systems in power reactors

International Nuclear Information System (INIS)

Aharon, J.; Harrari, R.; Weiss, Y.; Barnea, Y.; Katz, M.; Szanto, M.

1996-01-01

This paper reviews several R and D activities associated with the subject of passive cooling systems, conducted by the N.R.C.Negev thermohydraulic group. A short introduction considering different types of thermosyphons and their applications is followed by a detailed description of the experimental work, its results and conclusions. An ongoing research project is focused on the evaluation of the external dry air passive containment cooling system (PCCS) in the AP-600 (Westinghouse advanced pressurized water reactor). In this context some preliminary theoretical results and planned experimental research are for the fature described

ALWR safety approaches and trends. Implementation of passive safety features in the design

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V

1995-11-01

Reactor vendors world-wide are examining various advanced light water reactors (ALWR) options to reach utility goals. The amount of information available about each design varies essentially depending on its maturity. Some advanced reactor designs are the evolutionary results of combining old structures, systems and components in new ways, others use innovative solutions. A summary review is given for better understanding of new ALWR design trends and approaches in different countries and subsequent R and D activities. An attempt was made to describe and assess specific innovative and passive features implemented in the leading ALWR designs for further plant design safety improvements. The advantages and disadvantages of these innovations in obtaining reliable systems have been considered. Also, this report indicates the importance of uncertainties remaining and identifies the additional work needed. 51 refs, 27 figs, 7 tabs.

ALWR safety approaches and trends. Implementation of passive safety features in the design

International Nuclear Information System (INIS)

Ignatiev, V.

1995-11-01

Reactor vendors world-wide are examining various advanced light water reactors (ALWR) options to reach utility goals. The amount of information available about each design varies essentially depending on its maturity. Some advanced reactor designs are the evolutionary results of combining old structures, systems and components in new ways, others use innovative solutions. A summary review is given for better understanding of new ALWR design trends and approaches in different countries and subsequent R and D activities. An attempt was made to describe and assess specific innovative and passive features implemented in the leading ALWR designs for further plant design safety improvements. The advantages and disadvantages of these innovations in obtaining reliable systems have been considered. Also, this report indicates the importance of uncertainties remaining and identifies the additional work needed. 51 refs, 27 figs, 7 tabs

Preliminary design of an advanced programmable digital filter network for large passive acoustic ASW systems. [Parallel processor

Energy Technology Data Exchange (ETDEWEB)

McWilliams, T.; Widdoes, Jr., L. C.; Wood, L.

1976-09-30

The design of an extremely high performance programmable digital filter of novel architecture, the LLL Programmable Digital Filter, is described. The digital filter is a high-performance multiprocessor having general purpose applicability and high programmability; it is extremely cost effective either in a uniprocessor or a multiprocessor configuration. The architecture and instruction set of the individual processor was optimized with regard to the multiple processor configuration. The optimal structure of a parallel processing system was determined for addressing the specific Navy application centering on the advanced digital filtering of passive acoustic ASW data of the type obtained from the SOSUS net. 148 figures. (RWR)

Containment integrity analysis for the (W) advanced AP600

International Nuclear Information System (INIS)

Gagnon, A.F.; Howe, K.S.

1989-01-01

This paper reports that since 1987, Westinghouse has been performing containment cooling analyses in support of the Advanced AP600 plant design. This program was intended to verify the feasibility of the passive containment cooling system features of the AP600 design. To support this design, containment analyses of the AP600 containment for a large break LOCA and a large Steam Line Break were performed. The transient results indicate the feasibility of the passive containment design by demonstrating the capability to remove sufficient heat to limit containment atmosphere conditions to within acceptable limits following these postulated accidents. These results also indicate that the PCCS can reduce containment pressure to less than one-quarter design pressure at 24 hours following the most severe accident scenario thereby minimizing containment leakage concerns

Hood River Passive House

Energy Technology Data Exchange (ETDEWEB)

Hales, David [BA-PIRC, Spokane, WA (United States)

2014-01-01

The Hood River Passive Project was developed by Root Design Build of Hood River Oregon using the Passive House Planning Package (PHPP) to meet all of the requirements for certification under the European Passive House standards. The Passive House design approach has been gaining momentum among residential designers for custom homes and BEopt modeling indicates that these designs may actually exceed the goal of the U.S. Department of Energy's (DOE) Building America program to "reduce home energy use by 30%-50% (compared to 2009 energy codes for new homes). This report documents the short term test results of the Shift House and compares the results of PHPP and BEopt modeling of the project. The design includes high R-Value assemblies, extremely tight construction, high performance doors and windows, solar thermal DHW, heat recovery ventilation, moveable external shutters and a high performance ductless mini-split heat pump. Cost analysis indicates that many of the measures implemented in this project did not meet the BA standard for cost neutrality. The ductless mini-split heat pump, lighting and advanced air leakage control were the most cost effective measures. The future challenge will be to value engineer the performance levels indicated here in modeling using production based practices at a significantly lower cost.

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

International Nuclear Information System (INIS)

Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

2014-01-01

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

Regulatory Risk Management of Advanced Nuclear Power Plants

International Nuclear Information System (INIS)

George, Glenn R.

2002-01-01

Regulatory risk reflects both the likelihood of adverse outcomes during regulatory interactions and the severity of those outcomes. In the arena of advanced nuclear power plant licensing and construction, such adverse outcomes may include, for example, required design changes and construction delays. These, in turn, could significantly affect the economics of the plant and the generation portfolio in which it will operate. In this paper, the author addresses these issues through the lens of risk management. The paper considers various tools and techniques of regulatory risk management, including design diversity and hedging strategies. The effectiveness of alternate approaches is weighed and recommendations are made in several regulatory contexts. (author)

Investigations on passive containment cooling

International Nuclear Information System (INIS)

Knebel, J.U.; Cheng, X.; Neitzel, H.J.; Erbacher, F.J.; Hofmann, F.

1997-01-01

The composite containment design for advanced LWRs that has been examined under the PASCO project is a promising design concept for purely passive decay heat removal after a severe accident. The passive cooling processes applied are natural convection and radiative heat transfer. Heat transfer through the latter process removes at an emission coefficient of 0.9 about 50% of the total heat removed via the steel containment, and thus is an essential factor. The heat transferring surfaces must have a high emission coefficient. The sump cooling concept examined under the SUCO project achieves a steady, natural convection-driven flow from the heat source to the heat sink. (orig./CB) [de

Assessment of ALWR passive safety system reliability. Phase 1: Methodology development and component failure quantification

International Nuclear Information System (INIS)

Hake, T.M.; Heger, A.S.

1995-04-01

Many advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive systems to perform safety functions, rather than active systems as in current reactor designs. These passive systems depend to a great extent on physical processes such as natural circulation for their driving force, and not on active components, such as pumps. An NRC-sponsored study was begun at Sandia National Laboratories to develop and implement a methodology for evaluating ALWR passive system reliability in the context of probabilistic risk assessment (PRA). This report documents the first of three phases of this study, including methodology development, system-level qualitative analysis, and sequence-level component failure quantification. The methodology developed addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. Traditional PRA methods, such as fault and event tree modeling, are applied to the component failure aspect. Thermal-hydraulic calculations are incorporated into a formal expert judgment process to address uncertainties in selected natural processes and success criteria. The first phase of the program has emphasized the component failure element of passive system reliability, rather than the natural process uncertainties. Although cursory evaluation of the natural processes has been performed as part of Phase 1, detailed assessment of these processes will take place during Phases 2 and 3 of the program

A study on advanced man-machine interface system for autonomous nuclear power plants

International Nuclear Information System (INIS)

Matsuoka, Takeshi; Numano, Masayoshi; Fukuto, Junji; Sugasawa, Shinobu; Miyazaki, Keiko; Someya, Minoru; Haraki, Nobuo

1994-01-01

A man-machine interface(MMI) system of an autonomous nuclear power plant has an advanced function compared with that of the present nuclear power plants. The MMI has a function model of a plant state, and updates and revises this function model by itself. This paper describes the concept of autonomous nuclear power plants, a plant simulator of an autonomous power plant, a contracted function model of a plant state, three-dimensional color graphic display of a plant state, and an event-tree like expression for plant states. (author)

Recent Advances in Ocean Nuclear Power Plants

Directory of Open Access Journals (Sweden)

Kang-Heon Lee

2015-10-01

Full Text Available In this paper, recent advances in Ocean Nuclear Power Plants (ONPPs are reviewed, including their general arrangement, design parameters, and safety features. The development of ONPP concepts have continued due to initiatives taking place in France, Russia, South Korea, and the United States. Russia’s first floating nuclear power stations utilizing the PWR technology (KLT-40S and the spar-type offshore floating nuclear power plant designed by a research group in United States are considered herein. The APR1400 and SMART mounted Gravity Based Structure (GBS-type ONPPs proposed by a research group in South Korea are also considered. In addition, a submerged-type ONPP designed by DCNS of France is taken into account. Last, issues and challenges related to ONPPs are discussed and summarized.

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)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Prospects for advanced coal-fuelled fuel cell power plants

International Nuclear Information System (INIS)

Jansen, D.; Laag, P.C. van der; Oudhuis, A.B.J.; Ribberink, J.S.

1994-01-01

As part of ECN's in-house R and D programmes on clean energy conversion systems with high efficiencies and low emissions, system assessment studies have been carried out on coal gasification power plants integrated with high-temperature fuel cells (IGFC). The studies also included the potential to reduce CO 2 emissions, and to find possible ways for CO 2 extraction and sequestration. The development of this new type of clean coal technology for large-scale power generation is still far off. A significant market share is not envisaged before the year 2015. To assess the future market potential of coal-fuelled fuel cell power plants, the promise of this fuel cell technology was assessed against the performance and the development of current state-of-the-art large-scale power generation systems, namely the pulverized coal-fired power plants and the integrated coal gasification combined cycle (IGCC) power plants. With the anticipated progress in gas turbine and gas clean-up technology, coal-fuelled fuel cell power plants will have to face severe competition from advanced IGCC power plants, despite their higher efficiency. (orig.)

Implications of passive safety based on historical industrial experience

International Nuclear Information System (INIS)

Forsberg, C.W.

1988-01-01

In the past decade, there have been multiple proposals for applying different technologies to achieve passively safe light water reactors (LWRs). A key question for all such concepts is, ''What are the gains in safety, costs, and reliability for passive safety systems.'' Using several types of historical data, estimates have been made of gains from passive safety and operating systems, which are independent of technology. Proposals for passive safety in reactors usually have three characteristics: (1) Passive systems with no moving mechanical parts, (2) systems with far fewer components and (3) more stringent design criteria for safety-related and process systems. Each characteristic reduces the potential for an accident and may increase plant reliability. This paper addresses gains from items (1) and (2). Passive systems often allow adoption of more rigorous design criteria which would be either impossible or economically unfeasible for active systems. This important characteristic of passive safety systems cannot be easily addressed using historical industrial experience

Advanced safeguards research and development plan with an emphasis on its impact on nuclear power-plant design

International Nuclear Information System (INIS)

Tobin, S.J.; Demuth, S.F.; Miller, M.C.; Swinhoe, M.T.; Thomas, K.E.

2007-01-01

One tool for reducing the concern of nuclear proliferation is enhanced safeguards. Present safeguards have evolved over the past 40 years, and future safeguards will grow from this strong base to implement new technologies for improving our ability to quantify nuclear material. This paper will give an overview of the advanced technology research and development plan for safeguarding. One of the research facilities planned by the Department of Energy is the Advanced Fuel Cycle Facility (AFCF), to develop a novel nuclear fuel recycling program. Since the Advanced Fuel Cycle Facility will receive and reprocess spent fuel and will fabricate fast-reactor fuel, a wide breadth of safeguards technologies is involved. A fundamental concept in safeguards is material control and accounting (MCA). 4 topics concerning MCA and requiring further research have been identified: 1) measuring spent fuel, 2) measuring the plutonium content in the electro-refiner with pyro-processing, 3) measuring plutonium in the presence of other actinides, and 4) measuring neptunium and americium in the presence of other actinides. As for the long-term research and development plan for the AFCF, it will include improving MCA techniques as well as introducing new techniques that are not related to MCA, for example, enhanced containment and surveillance, or enhanced process monitoring. The top priority will stay quantifying the plutonium as accurately as possible and to reach this purpose 4 relevant technologies have been identified: 1) the microcalorimeter, 2) the passive neutron-albedo reactivity, 3) list-mode data acquisition, and 4) a liquid-scintillator multiplicity counter. Incorporating safeguards into the initial design of AFCF (safeguards by design) is a central concept. As the technology research and development plan for the Advanced Fuel Cycle Facility is examined, particular attention will be given to safeguards technologies that may affect the physical design of nuclear power plants

Simulation of a passive auxiliary feedwater system with TRACE5

Energy Technology Data Exchange (ETDEWEB)

Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo, E-mail: maloral@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), València (Spain)

2017-07-01

The study of the nuclear power plant accidents occurred in recent decades, as well as the probabilistic risk assessment carried out for this type of facility, present human error as one of the main contingency factors. For this reason, the design and development of generation III, III+ and IV reactors, which include inherent and passive safety systems, have been promoted. In this work, a TRACE5 model of ATLAS (Advanced Thermal- Hydraulic Test Loop for Accident Simulation) is used to reproduce an accidental scenario consisting in a prolonged Station BlackOut (SBO). In particular, the A1.2 test of the OECD-ATLAS project is analyzed, whose purpose is to study the primary system cooling by means of the water supply to one of the steam generators from a Passive Auxiliary Feedwater System (PAFS). This safety feature prevents the loss of secondary system inventory by means of the steam condensation and its recirculation. Thus, the conservation of a heat sink allows the natural circulation flow rate until restoring stable conditions. For the reproduction of the test, an ATLAS model has been adapted to the experiment conditions, and a PAFS has been incorporated. >From the simulation test results, the main thermal-hydraulic variables (pressure, flow rates, collapsed water level and temperature) are analyzed in the different circuits, contrasting them with experimental data series. As a conclusion, the work shows the TRACE5 code capability to correctly simulate the behavior of a passive feedwater system. (author)

Human factors survey of advanced instrumentation and controls technologies in nuclear plants

International Nuclear Information System (INIS)

Carter, R.J.

1992-01-01

A survey of advanced instrumentation and controls (I ampersand C) technologies and associated human factors issues in the US and Canadian nuclear industries was carried out by a team from Oak Ridge national laboratory to provide background for the development of regulatory policy, criteria, and guides for review of advanced I ampersand C systems as well as human engineering guidelines for evaluating these systems. The survey found those components of the US nuclear industry surveyed to be quite interested in advanced I ampersand C, but very cautious in implementing such systems in nuclear facilities and power plants. The trend in the facilities surveyed is to experiment cautiously when there is an intuitive advantage or short-term payoff. In the control room, the usual practice is direct substitution of digital and microprocessor-based instruments or systems that are functionally identical to the analog instruments or systems being replaced. The most advanced I ampersand C systems were found in the Canadian CANDU plants, where the newest plant has digital system in almost 100% of its control systems and in over 70% of its plant protection system. The hypothesis that properly 'introducing digital systems increases safety' is supported by the Canadian experience. The performance of these digital systems was achieved using an appropriate quality assurance program for the software development. The ability of digital systems to detect impending failures and initiate a fail-safe action, is a significant safety issue that should be of special interest to every US utility as well as to the US Nuclear Regulatory Commission. (orig.)

Main Steam Line Break Analysis for the Fully Passive Safety System of SMART

International Nuclear Information System (INIS)

Kim, Seong Wook; Chun, Ji Han; Bae, Kyoo Hwan; Kim, Keung Koo

2013-01-01

The standard design approval of SMART (System-integrated Modular Advanced ReacTor) developed by KAERI and KEPCO consortium was issued on July 4, 2012. Although SMART has enhanced safety compared to the conventional reactor, there is a demand to meet the 'passive safety performance requirements' after the Fukushima accident. The passive safety performance requirements are the capabilities to maintain the plant at a safe shutdown condition for a minimum of 72 hours without AC power supply or operator action in case of design basis accident (DBA). To satisfy the requirements, KAERI is developing a safety enhanced SMART by adopting a passive safety injection system. The passive safety injection system developed for SMART is a gravity-driven injection system, which consists of four trains, each of which includes a pressure balance line, core makeup tank (CMT), safety injection tank (SIT) and injection line. The CMT plays an important role to inject borated water into the RCS to prevent or dissolve the return to power (re-criticality) condition during the event of increase in heat removal by the secondary system. The main steam line break accident (MSLB) is the most limiting accident for an increase in heat removal by the secondary system. In this study, the safety analysis results of MSLBs at hot full power condition and at hot zero power condition in view of re-criticality are given. The MSLB accident has been analyzed for the SMART adopting fully passive safety system in the aspect of re-criticality. The results show that the core remains subcritical condition throughout the transient due to the borated water injected by the CMT. As further works, many kinds of analyses and sensitivity studies should be performed for the design establishment and improvement of the fully passive system of SMART

Application of passive auto catalytic recombiner (PAR) for BWR plants in Japan

International Nuclear Information System (INIS)

Kobayashi, K.; Murano, K.; Yamanari, S.; Yamamoto, Y.

2001-01-01

The passive auto-catalytic recombiner (PAR), which can recombine flammable gases such as hydrogen and oxygen with each other to avoid an explosion in case of a loss-of-coolant accident (LOCA), installed in the primary containment vessel does not require a power supply or dynamic equipment, while the existing flammability gas control system (FCS) of most BWRs as an outer loop of the primary containment vessel needs them to make flammable gases circulate through blowers and heaters in the system. PAR offers a number of advantages over existing FCS, such as high reliability, low cost due to much smaller amount of materials needed, good maintainability, good operability in case of a LOCA, and smaller space for installation. An experimental study has been carried out for the purpose of solving the problems of applying PAR to Japanese BWR plants instead of existing FCS, in which we grasped the basic characteristics of PAR. (author)

Neutralization of wastewater from nitrite passivation

International Nuclear Information System (INIS)

Pawlowski, L.; Mientki, B.; Wasag, H.

1982-01-01

A method for neutralization of wastewater formed in nitrite passivation has been presented. The method consists of introducing urea into wastewater and acidifying it with sulphuric acid. Wastewater is neutralized with lime. After clarification, wastewater can be drained outside the plant

The ARIES-AT advanced tokamak, Advanced technology fusion power plant

International Nuclear Information System (INIS)

Najmabadi, Farrokh; Abdou, A.; Bromberg, L.

2006-01-01

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N . ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

Advanced control and instrumentation systems in nuclear power plants. Design, verification and validation

International Nuclear Information System (INIS)

Haapanen, P.

1995-01-01

The Technical Committee Meeting on design, verification and validation of advanced control and instrumentation systems in nuclear power plants was held in Espoo, Finland on 20 - 23 June 1994. The meeting was organized by the International Atomic Energy Agency's (IAEA) International Working Group's (IWG) on Nuclear Power Plant Control and Instrumentation (NPPCI) and on Advanced Technologies for Water Cooled Reactors (ATWR). VTT Automation together with Imatran Voima Oy and Teollisuuden Voima Oy responded about the practical arrangements of the meeting. In total 96 participants from 21 countries and the Agency took part in the meeting and 34 full papers and 8 posters were presented. Following topics were covered in the papers: (1) experience with advanced and digital systems, (2) safety and reliability analysis, (3) advanced digital systems under development and implementation, (4) verification and validation methods and practices, (5) future development trends. (orig.)

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

Nuclear safety: operational aspects. 3. Hazard Analysis of Passive Systems

International Nuclear Information System (INIS)

Burgazzi, Luciano

2001-01-01

Interest has been aroused in recent years regarding the reliability assessment of passive systems being developed by suppliers, industries, utilities, and research organizations that aim at plant safety improvement and substantial simplification in its implementation. The approach to passive systems reliability assessment entails first a detailed system and safety analysis, and failure mode and effect analysis (FMEA) methodology has been chosen to perform the safety analysis at the system level. The FMEA technique allows identification of all potential failure modes in a system to evaluate their effects on the system and to classify them according to their severity; this technique identifies the reliability-critical areas in the system where modifications to the design are required to reduce the probability of failure. The present study concerns passive systems designed for decay heat removal relying on natural circulation that foresee, for the most part, a condenser immersed in a cooling pool. This is to identify and rank by importance the potential hazards related to passive-system equipment and operation that may critically affect the safety or availability of the plant. More specifically, the content of the paper analyzes the isolation condenser (IC) system foreseen for advanced boiling water reactors for removal of excess sensible and core decay heat by natural circulation during isolation transients. This FMEA analysis is the initial step to be accomplished as support for the development of a methodology aimed at the reliability assessment of thermal-hydraulic passive safety systems, providing important input to more detailed quantitative studies employing, for instance, event trees and fault trees or other reliability/availability models. Main purposes of the work are to identify important accident initiators, find out the possible consequences on the plant deriving from component failures, individuate possible causes, identify mitigating features and

CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

Energy Technology Data Exchange (ETDEWEB)

Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

2015-10-15

Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

Demonstration of Passive Fuel Cell Thermal Management Technology

Science.gov (United States)

Burke, Kenneth A.; Jakupca, Ian; Colozza, Anthony; Wynne, Robert; Miller, Michael; Meyer, Al; Smith, William

2012-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates and integrated heat exchanger technology to collect the heat from the cooling plates (Ref. 1). The next step in the development of this passive thermal approach was the demonstration of the control of the heat removal process and the demonstration of the passive thermal control technology in actual fuel cell stacks. Tests were run with a simulated fuel cell stack passive thermal management system outfitted with passive cooling plates, an integrated heat exchanger and two types of cooling flow control valves. The tests were run to demonstrate the controllability of the passive thermal control approach. Finally, successful demonstrations of passive thermal control technology were conducted with fuel cell stacks from two fuel cell stack vendors.

Engineered safeguards and passive safety features (safety analysis detailed report no. 6)

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The Safety-Analysis Summary lists the reactor's safety aspects for passive and active prevention of severe accidents and mitigation of accident consequences, i.e., intrinsic and passive protections of the plant; intrinsic and passive protections of the core; inherent decay-heat removal systems; rapid-shutdown systems; four physical containment barriers. This report goes into further details regarding some of this aspects.

Preservation of FFTF Data Related to Passive Safety Testing

International Nuclear Information System (INIS)

Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

2010-01-01

One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980's. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: (1) to verify natural circulation as a reliable means to safely remove decay heat, (2) to extend passive safety

A modeling and control approach to advanced nuclear power plants with gas turbines

International Nuclear Information System (INIS)

Ablay, Günyaz

2013-01-01

Highlights: • Load frequency control strategies in nuclear plants are researched. • Nuclear reactor-centered control system may not be suitable for load control. • Local unit controllers improve stability and overall time constant. • Coolant loops in nuclear plants should be controlled locally. - Abstract: Advanced nuclear power plants are currently being proposed with a number of various designs. However, there is a lack of modeling and control strategies to deal with load following operations. This research investigates a possible modeling approach and load following control strategy for gas turbine nuclear power plants in order to provide an assessment way to the concept designs. A load frequency control strategy and average temperature control mechanism are studied to get load following nuclear power plants. The suitability of the control strategies and concept designs are assessed through linear stability analysis methods. Numerical results are presented on an advanced molten salt reactor concept as an example nuclear power plant system to demonstrate the validity and effectiveness of the proposed modeling and load following control strategies

Advanced High Temperature Reactor Systems and Economic Analysis

Energy Technology Data Exchange (ETDEWEB)

Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

2011-09-01

The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

MARS, 600 MWth NUCLEAR POWER PLANT

International Nuclear Information System (INIS)

Cumo, M.; Naviglio, A.; Sorabella, L.

2004-01-01

MARS (Multipurpose Advanced Reactor, inherently Safe) is a 600 MWth, single loop, pressurized light water reactor (PWR), developed at the Dept. of Nuclear Engineering and Energy Conversion of the University of Rome ''La Sapienza''. The design was focused to a multipurpose reactor to be used in high population density areas also for industrial heat production and, in particular, for water desalting. Using the well-proven technology and the operation experience of PWRs, the project introduces a lot of innovative features hugely improving the safety performance while keeping the cost of KWh competitive with traditional large power plants. Extensive use of passive safety, in depth plant simplification and decommissioning oriented design were the guidelines along the design development. The latest development in the plant design, in the decommissioning aspects and in the experimental activities supporting the project are shown in this paper

European BWR R and D cluster for innovative passive safety systems

International Nuclear Information System (INIS)

Hicken, E.F.; Lensa, W. von

1996-01-01

The main technological innovation trends for future nuclear power plants tend towards a broader use of passive safety systems for the prevention, mitigation and managing of severe accident scenarios. Several approaches have been undertaken in a number of European countries to study and demonstrate the feasibility and charateristics of innovative passive safety systems. The European BWR R and D Cluster combines those experimental and analytical efforts that are mainly directed to the introduction of passive safety systems into boiling water reactor technology. The Cluster is grouped around thermohydraulic test facilities in Europe for the qualification of innovative BWR safety systems, also taking into account especially the operating experience of the nuclear power plant Dodewaard and other BWRs, which already incorporated some passive safety features. The background, the objectives, the structure of the project and the work programme are presented in this paper as well as an outline of the significance of the expected results. (orig.) [de

CMT scaling analysis and distortion evaluation in passive integral test facility

International Nuclear Information System (INIS)

Deng Chengcheng; Qin Benke; Wang Han; Chang Huajian

2013-01-01

Core makeup tank (CMT) is the crucial device of AP1000 passive core cooling system, and reasonable scaling analysis of CMT plays a key role in the design of passive integral test facilities. H2TS method was used to perform scaling analysis for both circulating mode and draining mode of CMT. And then, the similarity criteria for CMT important processes were applied in the CMT scaling design of the ACME (advanced core-cooling mechanism experiment) facility now being built in China. Furthermore, the scaling distortion results of CMT characteristic Ⅱ groups of ACME were calculated. At last, the reason of scaling distortion was analyzed and the distortion evaluation was conducted for ACME facility. The dominant processes of CMT circulating mode can be adequately simulated in the ACME facility, but the steam condensation process during CMT draining is not well preserved because the excessive CMT mass leads to more energy to be absorbed by cold metal. However, comprehensive analysis indicates that the ACME facility with high-pressure simulation scheme is able to properly represent CMT's important phenomena and processes of prototype nuclear plant. (authors)

Defence in Depth by Design for the Advanced GIII NPP in China

Energy Technology Data Exchange (ETDEWEB)

Liu, S.; Zhang, Y.; Zhang, X., E-mail: liusongtao.npic@gmail.com [Science and Technology on Reactor System Design Technology Laboratory Chengdu, Sichuan (China)

2014-10-15

This paper describes the design of the advanced nuclear power plant ACP1000 in China that keeps the principle of defence in depth. To enhance the safety of the new generation NPPs, passive and active engineering safety features are used. The reactor will be kept safe under design basis accidents by using active engineering safety features, such as the medium and low pressure safety injection systems, and the emergency feedwater system. Under beyond DBAs, the passive safety systems will be actuated to keep removing residual heat for more than 72 hours, and to keep the core melt retained and cooled in the vessel. After the Fukushima nuclear accident, there are six main design enhancements in ACP1000 to meet the demands of the China authorities. (author)

Preliminary investigation on reliability assessment of passive safety system

International Nuclear Information System (INIS)

Huang Changfan; Kuang Bo

2012-01-01

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

Passive containment system for a nuclear reactor

International Nuclear Information System (INIS)

Kleimola, F.W.

1976-01-01

A containment system is described that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means

ESBWR-an economical passive plant design

International Nuclear Information System (INIS)

Arnold, H.; Stoop, P.M.; Gonzales, A.; Rao, A.

1996-01-01

The ESBWR is a plant design that builds on the GKN Dodewaard natural-circulation reactor and the simplified boiling water reactor (SBWR) design. The major objective of the ESBWR program, which has been in place for the past 3 yr, is to develop a plant design with proven technology that improves the overall plant economics. It utilizes the experience and basic simplicity of the Dodewaard plant and 670-MW(electric) SBWR design features. The design is being developed by an international team of utilities, designers, and researchers. It is being designed to meet the utility and regulatory requirements of Europe. It also addresses the key economic challenges for future nuclear power stations

Natural circulation in water cooled nuclear power plants: Phenomena, models, and methodology for system reliability assessments

International Nuclear Information System (INIS)

2005-11-01

In recent years it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. Further, the IAEA Conference on The Safety of Nuclear Power: Strategy for the Future which was convened in 1991 noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to assure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are an ongoing activity in several IAEA Member States. Some new designs also utilize natural circulation as a means to remove core power during normal operation. In response to the motivating factors discussed above, and to foster international collaboration on the enabling technology of passive systems that utilize natural circulation, an IAEA Coordinated Research Project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation was started in early 2004. Building on the shared expertise within the CRP, this publication presents extensive information on natural circulation phenomena, models, predictive tools and experiments that currently support design and analyses of natural circulation systems and highlights areas where additional research is needed. Therefore, this publication serves both to provide a description of the present state of knowledge on natural circulation in water cooled nuclear power plants and to guide the planning and conduct of the CRP in

Nuclear heat source design for an advanced HTGR process heat plant

International Nuclear Information System (INIS)

McDonald, C.F.; O'Hanlon, T.W.

1983-01-01

A high-temperature gas-cooled reactor (HTGR) coupled with a chemical process facility could produce synthetic fuels (i.e., oil, gasoline, aviation fuel, methanol, hydrogen, etc.) in the long term using low-grade carbon sources (e.g., coal, oil shale, etc.). The ultimate high-temperature capability of an advanced HTGR variant is being studied for nuclear process heat. This paper discusses a process heat plant with a 2240-MW(t) nuclear heat source, a reactor outlet temperature of 950 0 C, and a direct reforming process. The nuclear heat source outputs principally hydrogen-rich synthesis gas that can be used as a feedstock for synthetic fuel production. This paper emphasizes the design of the nuclear heat source and discusses the major components and a deployment strategy to realize an advanced HTGR process heat plant concept

MITI project on advanced man-machine system for nuclear power plants

International Nuclear Information System (INIS)

Kato, Kanji; Watanabe, Takao; Hayakawa, Hiroyasu; Naito, Norio; Masui, Takao; Ogino, Takamichi.

1988-01-01

A computerized operator support system (COSS) against abnormal plant conditions was developed as a five-year project from 1980 to 1984, under the sponsorship of the Ministry of International Trade and Industry. The main purpose of the COSS development was to implement the lessons learned from the Three Mile Island accident. The main nuclear industries in Japan participated in the project. The design concept of the operator support functions and the method to implement it were established, and the prototype systems of the COSS for BWR and PWR plants were developed. After the completion of the COSS development, the above participant group once again joined for the work on an advanced man-machine system for nuclear power plants (MMS-NPP). This eight-year project aims at realizing an advanced operator support system by applying artificial intelligence, especially knowledge engineering, and sophisticated man-machine interface devices. Its main objectives are shown. This system configuration, operating method decision system, man-machine communication system, operation and maintenance support functions and so on are described. (Kako, I.)

Conceptual study of advanced PWR systems. A study of passive and inherent safety design concepts for advanced light water reactors

Energy Technology Data Exchange (ETDEWEB)

Chang, Soon Heung; No, Hee Cheon; Baek, Won Pil; Jae, Shim Young; Lee, Goung Jin; Na, Man Gyun; Lee, Jae Young; Kim, Han Gon; Kang, Ki Sig; Moon, Sang Ki; Kim, Yun Il; Park, Jae Wook; Yang, Soo Hyung; Kim, Soo Hyung; Lee, Seong Wook; Kim, Hong Che; Park, Hyun Sik; Jeong, Ji Hwan; Lee, Sang Il; Jung, Hae Yong; Kim, Hyong Tae; Chae, Kyung Sun; Moon, Ki Hoon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1995-08-01

The five thermal-hydraulic concepts chosen for advanced PWR have been studied as follows: (1) Critical Heat Flux: Review of previous works, analysis of parametric trends, analysis of transient CHF characteristics, extension of the CHF date bank, survey and assessment of correlations, design of a intermediate-pressure CHF test loop have been performed. (2) Passive Cooling Concepts for Concrete Containment system: Review of condensation phenomena with noncondensable gases, selection of a promising concept (i.e., use of external condensers), design of test loop according to scaling laws have been accomplished. and computer programs based on the control-volume approach, and the conceptual design of test loop have been accomplished. (4) Fluidic Diode Concepts: Review of previous applications of the concept, analysis major parameters affecting the performance, development of a computational code, and conceptual investigation of the verification test loop have been performed. (5) Wet Thermal Insulator: Review of previous works, selection of promising methods ( i.e. ceramic fiber in a steel case and mirror-type insulator), and conceptual design of the experimental loop have been performed. (author). 9 refs.

Proceedings of 2009 international congress on advances in nuclear power plants

International Nuclear Information System (INIS)

2009-01-01

This CD-ROM is the collection of the paper presented at the 2009 International Congress on Advances in Nuclear Power Plants (ICAPP'09) . The 365 of the presented papers are indexed individually. (J.P.N.)

Active versus passive screening for entrance control

International Nuclear Information System (INIS)

McCormick, N.J.

1976-01-01

The benefits of different entrance control actions are quantitatively assessed by defining a relative improvement index for the screening activity. Three classes of entrance control measures are investigated: the use of a purely active screening measure (such as a portal monitor), the use of a purely passive screening measure (such as personality typing), and the combined use of active and passive measures. Active entrance control measures have been studied previously [McCormick and Erdmann, Nucl. Mat. Manag. 4, (1975)] where it was determined that the relative improvement index is approximately related to the nondetection probability factor r for the protective system by (1-r + r ln r). It is shown here that the relative improvement index for a purely passive screening system also can be approximately expressed in a convenient manner. Because the probability is very small that a sabotage or diversion action would be attempted, the result for passive screening, multiplied by r, may be combined with the factor (1-r + r ln r) to give the relative improvement index for a combined, active-and-passive entrance control system. Results from simple example calculations indicate that passive screening of nuclear plant personnel or applicants for such positions is orders-of-magnitude less effective than portal monitors or reasonable improvements in them. 5 tables

An approach for assessing ALWR passive safety system reliability

International Nuclear Information System (INIS)

Hake, T.M.

1991-01-01

Many advanced light water reactor designs incorporate passive rather than active safety features for front-line accident response. A method for evaluating the reliability of these passive systems in the context of probabilistic risk assessment has been developed at Sandia National Laboratories. This method addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. These processes provide the system's driving force; examples are natural circulation and gravity-induced injection. This paper describes the method, and provides some preliminary results of application of the approach to the Westinghouse AP600 design

Chemical monitoring strategy for the assessment of advanced water treatment plant performance.

Science.gov (United States)

Drewes, J E; McDonald, J A; Trinh, T; Storey, M V; Khan, S J

2011-01-01

A pilot-scale plant was employed to validate the performance of a proposed full-scale advanced water treatment plant (AWTP) in Sydney, Australia. The primary aim of this study was to develop a chemical monitoring program that can demonstrate proper plant operation resulting in the removal of priority chemical constituents in the product water. The feed water quality to the pilot plant was tertiary-treated effluent from a wastewater treatment plant. The unit processes of the AWTP were comprised of an integrated membrane system (ultrafiltration, reverse osmosis) followed by final chlorination generating a water quality that does not present a source of human or environmental health concern. The chemical monitoring program was undertaken over 6 weeks during pilot plant operation and involved the quantitative analysis of pharmaceuticals and personal care products, steroidal hormones, industrial chemicals, pesticides, N-nitrosamines and halomethanes. The first phase consisted of baseline monitoring of target compounds to quantify influent concentrations in feed waters to the plant. This was followed by a period of validation monitoring utilising indicator chemicals and surrogate measures suitable to assess proper process performance at various stages of the AWTP. This effort was supported by challenge testing experiments to further validate removal of a series of indicator chemicals by reverse osmosis. This pilot-scale study demonstrated a simplified analytical approach that can be employed to assure proper operation of advanced water treatment processes and the absence of trace organic chemicals.

New technologies deployment for advanced power plants

International Nuclear Information System (INIS)

Kiyoshi, Yamauchi

2007-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has been the total engineering and manufacturing company of pressurized water reactors (PWRs) in Japan since the commencement of commercial operations of Mihama Unit 1 of the Kansai Electric Power Company in 1970. Over these decades, MHI has endeavored to develop a broad spread of nuclear technology, from design, manufacturing, and construction, to plant maintenance services. More recently, with the ever rising need for nuclear power generation around the world to prevent global warming and to cope with surging oil prices, MHI is striving to expand its nuclear power business in the world market, such as US-APWR (Advanced Pressurized Water Reactor) in the U.S., as well as to develop technology for advanced reactors and nuclear fuel cycles to ensure energy security in the future. This paper introduces these approaches, especially focused on new technologies deployment for the global needs, and clarifies the current status and future prospects of MHI as the world's leading nuclear company. (author)

Advanced I and C systems for nuclear power plants feedback of experience

International Nuclear Information System (INIS)

Prehler, H.J.

2001-01-01

Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS.(author)

Technology Solutions Case Study: Evaluation of Passive Vents in New-Construction Multifamily Buildings

Energy Technology Data Exchange (ETDEWEB)

S. Puttagunta, S. Maxwell, D. Berger, and M. Zuluaga

2015-10-01

The Consortium for Advanced Residential Buildings (CARB) conducted research to gain more insight into passive vents. Because passive vents are meant to operate in a general environment of negative apartment pressure, the research assessed whether these negative pressures prevail through a variety of environmental conditions.

Advanced digital technology - improving nuclear power plant performance through maintainability

International Nuclear Information System (INIS)

Ford, J.L.; Senechal, R.R.; Altenhein, G.D.; Harvey, R.P.

1998-01-01

In today's energy sector there is ever increasing pressure on utilities to operate power plants at high capacity factors. To ensure nuclear power is competitive into the next century, it is imperative that strategic design improvements be made to enhance the performance of nuclear power plants. There are a number of factors that affect a nuclear power plant's performance; lifetime maintenance is one of the major contributors. The maturing of digital technology has afforded ABB the opportunity to make significant design improvements in the area of maintainability. In keeping with ABB's evolutionary advanced nuclear plant design approach, digital technology has systematically been incorporated into the control and protection systems of the most recent Korean nuclear units in operation and under construction. One example of this was the multi-functional design team approach that was utilized for the development of ABB's Digital Plant Protection System (DPPS). The design team consisted of engineers, maintenance technicians, procurement specialists and manufacturing personnel in order to provide a complete perspective on all facets of the design. The governing design goals of increased reliability and safety, simplicity of design, use of off-the-shelf products and reduced need for periodic surveillance testing were met with the selection of proven ABB-Advant Programmable Logic Controllers (PLCs) as the heart of the DPPS. The application of digital PLC technology allows operation for extended periods without requiring routine maintenance or re-calibration. A well documented commercial dedication program approved by the United States Nuclear Regulatory Commission (US NRC) as part of the System 80+ TM Advanced Light Water Reactor Design Certification Program, allowed the use of off-the shelf products in the design of the safety protection system. In addition, a number of mechanical and electrical improvements were made which support maintainability. The result is a DPPS

Passivation Strategies on Board Airbus ds Leo Pcdus

Directory of Open Access Journals (Sweden)

Lapeña Emilio

2017-01-01

This paper deals with the different strategies followed in the Airbus DS LEO PCDUs regarding the implementation of the passivation function in several LEO missions with different architectures (DET and MPPT solar array power conditioning. In the selection of the solution implemented in the frame of every mission, a key driver is the degree of advance in the test performed over flight representative battery modules regarding their safe behavior when deeply depleted after a long period in orbit with the passivation applied over the spacecraft.

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

Energy Technology Data Exchange (ETDEWEB)

Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

2006-12-11

This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

Scaling for Mixed Convection Heat Transfer in Passive Containments and Experiment Design

International Nuclear Information System (INIS)

Wang, Shengfei; Yu, Yu; Lv, Xuefeng; Niu, Fenglei; Yan, Xiuping

2012-01-01

Most of the advanced nuclear reactor design utilizes passive systems to remove heat from the core by natural circulation. The passive systems will be widely used in generation III pressurized water reactor. One of the typical passive systems is passive containment cooling system (PCCS), which is a passive condenser system designed to remove heat from the containment for long term cooling after a postulated reactor accident. In order to establish empirical correlations and develop simulation models, a scaling analysis is performed in designing an experiment for the prototype PCCS. This paper presents a scaling method and the design of the experimental facility. The key dimensionless parameters governing the dominant processes are given at last

Recent development of advanced BWR technology for plant application

International Nuclear Information System (INIS)

Horiuchi, Tetsuo; Sakurai, Mikio; Mase, Noriaki; Oyamada, Osamu; Nakadaira, Shiro.

1988-01-01

The development of advanced BWRs (ABWR) was completed in 1985. Through the authorization as the third improved and standardized plants of LWRs by the Ministry of International Trade and Industry, the detailed design phase of the actual project has begun, and the improved technology to be applied to the plants has been steadily verified and put in practical use. In the ABWRs, the operational capability, safety and economical efficiency as the general characteristics of the plants were further heightened by simplifying, heightening the performance and compactifying. Particularly internal pumps brought about the improvement of the operational capability and safety of the plants together with improved control rod driving system, besides promoting the simplification and compactification of the reactor system. Also the reinforced concrete containment vessels constructed into one body with the buildings have the compact structure and the solid property sufficiently withstanding hypothetical accidents, and contribute to the improvement of the economical efficiency and safety. These key improvement technologies completed their tests for practical use, and it was shown that the expected objectives are realized with the characteristics of high level, thus the steady steps toward the construction of the actual plants are promoted. (Kako, I.)

Incorporation of passive components aging into PRAs

International Nuclear Information System (INIS)

Phillips, J.H.; Roesener, W.S.; Magleby, H.L.; Geidl, V.

1993-01-01

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants. (orig.)

High-Throughput Phenotyping of Wheat and Barley Plants Grown in Single or Few Rows in Small Plots Using Active and Passive Spectral Proximal Sensing

OpenAIRE

Barmeier, Gero;Schmidhalter, Urs

2017-01-01

In the early stages of plant breeding, breeders evaluate a large number of varieties. Due to limited availability of seeds and space, plot sizes may range from one to four rows. Spectral proximal sensors can be used in place of labour-intensive methods to estimate specific plant traits. The aim of this study was to test the performance of active and passive sensing to assess single and multiple rows in a breeding nursery. A field trial with single cultivars of winter barley and winter wheat w...

Development activities on advanced LWR in Argentina

International Nuclear Information System (INIS)

Gomez, S.E.

2001-01-01

CAREM, an Argentinean project, consists of the development, design and construction of a small Nuclear Power Plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors in the world. The CAREM is an indirect cycle reactor with some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary cooling system, self-pressurized, primary cooling by natural circulation and safety system relying on passive features. In this paper a brief description of the CAREM distinctive features and associated development activities are presented. (author)

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)

Advanced exergoenvironmental analysis of a near-zero emission power plant with chemical looping combustion.

Science.gov (United States)

Petrakopoulou, Fontina; Tsatsaronis, George; Morosuk, Tatiana

2012-03-06

Carbon capture and storage (CCS) from power plants can be used to mitigate CO(2) emissions from the combustion of fossil fuels. However, CCS technologies are energy intensive, decreasing the operating efficiency of a plant and increasing its costs. Recently developed advanced exergy-based analyses can uncover the potential for improvement of complex energy conversion systems, as well as qualify and quantify plant component interactions. In this paper, an advanced exergoenvironmental analysis is used for the first time as means to evaluate an oxy-fuel power plant with CO(2) capture. The environmental impacts of each component are split into avoidable/unavoidable and endogenous/exogenous parts. In an effort to minimize the environmental impact of the plant operation, we focus on the avoidable part of the impact (which is also split into endogenous and exogenous parts) and we seek ways to decrease it. The results of the advanced exergoenvironmental analysis show that the majority of the environmental impact related to the exergy destruction of individual components is unavoidable and endogenous. Thus, the improvement potential is rather limited, and the interactions of the components are of lower importance. The environmental impact of construction of the components is found to be significantly lower than that associated with their operation; therefore, our suggestions for improvement focus on measures concerning the reduction of exergy destruction and pollutant formation.

Development and validation of advanced oxidation protective coatings for super critical steam power plant

Energy Technology Data Exchange (ETDEWEB)

Henderson, M.B.; Scheefer, M. [Alstom Power Ltd., Rugby (United Kingdom); Agueero, A. [Instituto Nacional de Tecnica Aerospacial (INTA) (Spain); Allcock, B. [Monitor Coatings Ltd. (United Kingdom); Norton, B. [Indestructible Paints Ltd. (United Kingdom); Tsipas, D.N. [Aristotle Univ. of Thessaloniki (Greece); Durham, R. [FZ Juelich (Germany); Xiang, Z. [Northumbria Univ. (United Kingdom)

2006-07-01

Increasing the efficiency of coal-fired power plant by increasing steam temperatures and pressures brings benefits in terms of cheaper electricity and reduced emissions, particularly CO{sub 2}. In recent years the development of advanced 9%Cr ferritic steels with improved creep strength has enabled power plant operation at temperatures in excess of 600 C, such that these materials are being exploited to construct a new generation of advanced coalfired plant. However, the move to higher temperatures and pressures creates an extremely hostile oxidising environment. To enable the full potential of the new steels to be achieved, it is vital that protective coatings are developed, validated under high temperature steam and applied to candidate components from the steam path. This paper reviews recent work conducted within the Framework V project ''Coatings for Supercritical Steam Cycles'' (SUPERCOAT) to develop and demonstrate advanced slurry and thermal spray coatings capable of providing steam oxidation protection at temperatures in excess of 620 C and up to 300 bar. The programme of work has demonstrated the feasibility of applying a number of candidate coatings to steam turbine power plant components and has generated long-term steam oxidation rate and failure data that underpin the design and application work packages needed to develop and establish this technology for new and retrofit plant. (orig.)

Improvement potential of a real geothermal power plant using advanced exergy analysis

International Nuclear Information System (INIS)

Gökgedik, Harun; Yürüsoy, Muhammet; Keçebaş, Ali

2016-01-01

The main purpose of this paper is to quantitatively evaluate thermodynamic performance of a geothermal power plant (GPP) from potential for improvement point of view. Thus, sources of inefficiency and irreversibilities can be determined through exergy analysis. The advanced exergy analysis is more appropriate to determine real potential for thermodynamic improvements of the system by splitting exergy destruction into unavoidable and avoidable portions. The performance critical components and the potential for exergy efficiency improvement of a GPP were determined by means of the advanced exergy analysis. This plant is the Bereket GPP in Denizli/Turkey as a current operating system. The results show that the avoidable portion of exergy destruction in all components except for the turbines is higher than the unavoidable value. Therefore, much can be made to lessen the irreversibilities for components of the Bereket GPP. The total exergy efficiency of the system is found to be 9.60%. Its efficiency can be increased up to 15.40% by making improvements in the overall components. Although the heat exchangers had lower exergy and modified exergy efficiencies, their exergy improvement potentials were high. Finally, in the plant, the old technology is believed to be one of the main reasons for low efficiencies. - Highlights: • Evaluation of potential for improvement of a GPP using advanced exergy analysis. • Efficiency can be increased up to 15.40% by making improvements in the components. • Heat exchangers are the highest avoidable values, making them the least efficient components in plant. • The main reasons for low efficiencies are believed to be the old technology.

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

Energy Technology Data Exchange (ETDEWEB)

Wood, RT

2004-09-27

This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

National project : advanced robot for nuclear power plant

International Nuclear Information System (INIS)

Tsunemi, T.; Takehara, K.; Hayashi, T.; Okano, H.; Sugiyama, S.

1993-01-01

The national project 'Advanced Robot' has been promoted by the Agency of Industrial science and Technology, MITI for eight years since 1983. The robot for a nuclear plant is one of the projects, and is a prototype intelligent one that also has a three dimensional vision system to generate an environmental model, a quadrupedal walking mechanism to work on stairs and four fingered manipulators to disassemble a valve with a hand tool. Many basic technologies such as an actuator, a tactile sensor, autonomous control and so on progress to high level. The prototype robot succeeded functionally in official demonstration in 1990. More refining such as downsizing and higher intelligence is necessary to realize a commercial robot, while basic technologies are useful to improve conventional robots and systems. This paper presents application studies on the advanced robot technologies. (author)

Safety significance of ATR passive safety response attributes

International Nuclear Information System (INIS)

Atkinson, S.A.

1990-01-01

The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety 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 to 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) models and 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 firewater injection system (emergency coolant system)

Advanced Light Water Reactor Plants System 80+trademark Design Certification Program

International Nuclear Information System (INIS)

Davis, G.A.

1992-01-01

Since 1985, ABB Combustion Engineering Nuclear Power (CENP) and Duke Engineering ampersand Services, Inc. (DE ampersand S) have been developing the next generation of pressurized water reactor (PWR) plant for worldwide deployment. The goal is to make available a pre-licensed, standardized plant design that can satisfy the need for a reliable and economic supply of electricity for residential, commercial and industrial use. To ensure that such a design is available when needed, it must be based on proven technology and established licensing criteria. These requirements dictate development of nuclear technology that is advanced, yet evolutionary in nature. This has been achieved with the System 80+ Standard Plant Design

Simulation-based biagnostics and control for nuclar power plants

International Nuclear Information System (INIS)

Lee, J.C.

1993-01-01

Advanced simulation-based diagnostics and control guidance systems for the identification and management of off-normal transient events in nuclear power plants is currently under investigation. To date a great deal of progress has been made in effectively and efficiently combining information obtained through fuzzy pattern recognition and macroscopic mass and energy inventory analysis for use in multiple failure diagnostics. Work has also begun on the unique problem of diagnostics and surveillance methodologies for advanced passively-safe reactors systems utilizing both statistical and fuzzy information. Plans are also being formulated for the development of deterministic optimal control algorithms combined with Monte Carlo incremental learning algorithms to be used for the flexible and efficient control of reactor transients

Piping reliability improvement through passive seismic supports

International Nuclear Information System (INIS)

Baltus, R.; Rubbers, A.

1999-01-01

The nuclear plants designed in the 1970's were equipped with large quantities of snubbers in auxiliary piping systems. The experience revealed a poor performance of snubbers during periodic inspection, while non-nuclear facility piping survived through strong earthquakes. Consequently, seismic design rules evolved towards more realistic criteria and passive dynamic supports were developed to reduce snubber quantities. These solutions improve the pipe reliability during normal operation while reducing the radiation exposure in a sample line is presented with the impact on pipe stresses compared to the results obtained with passive supports named Limit Stops. (author)

Advanced applications of water cooled nuclear power plants

International Nuclear Information System (INIS)

2008-07-01

By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average nuclear plant availability during 2006 reached 83%; - Growing energy needs: All forecasts project increases in world energy demand, especially as population and economic productivity grow. The strategies are country dependent, but usually involve a mix of energy sources; - Interest in advanced applications of nuclear energy, such as seawater desalination, steam for heavy oil recovery and heat and electricity for hydrogen production; - Environmental concerns and constraints: The Kyoto Protocol has been in force since February 2005, and for many countries (most OECD countries, the Russian Federation, the Baltics and some countries of the Former Soviet Union and Eastern Europe) greenhouse gas emission limits are imposed; - Security of energy supply is a national priority in essentially every country; and - Nuclear power is economically competitive and provides stability of electricity price. In the near term most new nuclear plants will be evolutionary water cooled reactors (Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs), often pursuing economies of scale. In the longer term, innovative designs that

Introduction to the 'CAS' nuclear propulsion plant for ships: specific safety options

International Nuclear Information System (INIS)

Verdeau, J.J.; Baujat, J.

1978-01-01

After a brief review of the development of nuclear propulsion in FRANCE (Land Based Prototype PAT 1964 - Navy nuclear ships - Advanced Nuclear Boiler Prototype CAP 1975 and now the CAS nuclear plant), the specific safety options of CAS are presented: cold, compartmented fuel (plates); reduced flow during LOCA; permanent cooling of fuel during LOCA; pressurized, entirely passive containment; no control rod ejection and possibility of temporary storage of spent fuel on board [fr

Passive BWR integral LOCA testing at the Karlstein test facility INKA

Energy Technology Data Exchange (ETDEWEB)

Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

2014-05-15

KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

Underwater Photosynthesis of Submerged Plants – Recent Advances and Methods

Science.gov (United States)

Pedersen, Ole; Colmer, Timothy D.; Sand-Jensen, Kaj

2013-01-01

We describe the general background and the recent advances in research on underwater photosynthesis of leaf segments, whole communities, and plant dominated aquatic ecosystems and present contemporary methods tailor made to quantify photosynthesis and carbon fixation under water. The majority of studies of aquatic photosynthesis have been carried out with detached leaves or thalli and this selectiveness influences the perception of the regulation of aquatic photosynthesis. We thus recommend assessing the influence of inorganic carbon and temperature on natural aquatic communities of variable density in addition to studying detached leaves in the scenarios of rising CO2 and temperature. Moreover, a growing number of researchers are interested in tolerance of terrestrial plants during flooding as torrential rains sometimes result in overland floods that inundate terrestrial plants. We propose to undertake studies to elucidate the importance of leaf acclimation of terrestrial plants to facilitate gas exchange and light utilization under water as these acclimations influence underwater photosynthesis as well as internal aeration of plant tissues during submergence. PMID:23734154

Research and development of advanced robots for nuclear power plants

International Nuclear Information System (INIS)

Tsukune, Hideo; Hirukawa, Hirohisa; Kitagaki, Kosei; Liu, Yunhui; Onda, Hiromu; Nakamura, Akira

1994-01-01

Social and economic demands have been pressing for automation of inspection tasks, maintenance and repair jobs of nuclear power plants, which are carried out by human workers under circumstances with high radiation level. Since the plants are not always designed for introduction of automatic machinery, sophisticated robots shall play a crucial role to free workers from hostile environments. We have been studying intelligent robot systems and regarded nuclear industries as one of the important application fields where we can validate the feasibility of the methods and systems we have developed. In this paper we firstly discuss on the tasks required in nuclear power plants. Secondly we introduce current status of R and D on special purpose robots, versatile robots and intelligent robots for automatizing the tasks. Then we focus our discussions on three major functions in realizing robotized assembly tasks under such unstructured environments as in nuclear power plants; planning, vision and manipulation. Finally we depict an image of a prototype robot system for nuclear power plants based on the advanced functions. (author) 64 refs

A Review on Passive and Integrated Near-Field Microwave Biosensors

Science.gov (United States)

Guha, Subhajit; Jamal, Farabi Ibne

2017-01-01

In this paper we review the advancement of passive and integrated microwave biosensors. The interaction of microwave with biological material is discussed in this paper. Passive microwave biosensors are microwave structures, which are fabricated on a substrate and are used for sensing biological materials. On the other hand, integrated biosensors are microwave structures fabricated in standard semiconductor technology platform (CMOS or BiCMOS). The CMOS or BiCMOS sensor technology offers a more compact sensing approach which has the potential in the future for point of care testing systems. Various applications of the passive and the integrated sensors have been discussed in this review paper. PMID:28946617

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

Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

International Nuclear Information System (INIS)

Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

1996-01-01

The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Passivation layer breakdown during laser-fired contact formation for photovoltaic devices

International Nuclear Information System (INIS)

Raghavan, A.; DebRoy, T.; Palmer, T. A.

2014-01-01

Low resistance laser-fired ohmic contacts (LFCs) can be formed on the backside of Si-based solar cells using microsecond pulses. However, the impact of these longer pulse durations on the dielectric passivation layer is not clear. Retention of the passivation layer during processing is critical to ensure low recombination rates of electron-hole pairs at the rear surface of the device. In this work, advanced characterization tools are used to demonstrate that although the SiO 2 passivation layer melts directly below the laser, it is well preserved outside the immediate LFC region over a wide range of processing parameters. As a result, low recombination rates at the passivation layer/wafer interface can be expected despite higher energy densities associated with these pulse durations.

Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-09-15

during LTO. It should be pointed out here that LTO has different meanings in different countries. For example, in the United States of America, LTO refers to operation beyond the original 40 year licence period. That is, a nuclear plant in the USA can add 20 years to its licensed length of operation, extending the plant life to 60, 80, or more years in 20 year increments. In other countries such as Japan, LTO refers to operations beyond 30 years; while advanced gas cooled reactors (AGRs) in the United Kingdom may extend their licensed life by five years at a time beyond the original 30 years of licensed length. One may divide the SSCs of a nuclear plant into two general classes: those that are active components, such as pumps, motors, turbogenerators, valves, compressors, sensors and actuators, and those that are passive components, such as the reactor vessel, piping, reactor internals, containment structure, cables and the like. For active components (e.g. rotating machinery), there are plenty of SDP techniques, with the exception of prognostics, that are proven and routinely used. The advances in this area have occurred in the ability to see the degradation more quickly and more clearly through the use of high resolution data and improved data processing and visualization techniques. The same is not true for passive components. For passive components, periodic in-service inspections (ISIs) are implemented in accordance with ageing management plans, using non-destructive examination (NDE) techniques, such as eddy current testing and ultrasonic wave measurements. These measurements are defined in numerous codes and standards that have been available and used for years, not only in the nuclear industry but also in aerospace and other fields. While effective, the NDE techniques do not normally provide in situ, continuous on-line, or remote testing capabilities.

Advanced Surveillance, Diagnostic and Prognostic Techniques in Monitoring Structures, Systems and Components in Nuclear Power Plants

International Nuclear Information System (INIS)

2013-01-01

during LTO. It should be pointed out here that LTO has different meanings in different countries. For example, in the United States of America, LTO refers to operation beyond the original 40 year licence period. That is, a nuclear plant in the USA can add 20 years to its licensed length of operation, extending the plant life to 60, 80, or more years in 20 year increments. In other countries such as Japan, LTO refers to operations beyond 30 years; while advanced gas cooled reactors (AGRs) in the United Kingdom may extend their licensed life by five years at a time beyond the original 30 years of licensed length. One may divide the SSCs of a nuclear plant into two general classes: those that are active components, such as pumps, motors, turbogenerators, valves, compressors, sensors and actuators, and those that are passive components, such as the reactor vessel, piping, reactor internals, containment structure, cables and the like. For active components (e.g. rotating machinery), there are plenty of SDP techniques, with the exception of prognostics, that are proven and routinely used. The advances in this area have occurred in the ability to see the degradation more quickly and more clearly through the use of high resolution data and improved data processing and visualization techniques. The same is not true for passive components. For passive components, periodic in-service inspections (ISIs) are implemented in accordance with ageing management plans, using non-destructive examination (NDE) techniques, such as eddy current testing and ultrasonic wave measurements. These measurements are defined in numerous codes and standards that have been available and used for years, not only in the nuclear industry but also in aerospace and other fields. While effective, the NDE techniques do not normally provide in situ, continuous on-line, or remote testing capabilities

Westinghouse Small Modular Reactor passive safety system response to postulated events

International Nuclear Information System (INIS)

Smith, M. C.; Wright, R. F.

2012-01-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 R reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

An experimental investigation of natural circulated air flow in the passive containment cooling system

International Nuclear Information System (INIS)

Ryu, S.H.; Oh, S.M.; Park, G.C.

2004-01-01

The objective of this study is to investigate the effects of air inlet position and external conditions on the natural circulated air flow rate in a passive containment cooling system of the advanced passive reactor. Experiments have been performed with 1/36 scaled segment type passive containment test facility. The air velocities and temperatures are measured through the air flow path. Also, the experimental results are compared with numerical calculations and show good agreement. (author)

Advanced control systems to improve nuclear power plant reliability and efficiency

International Nuclear Information System (INIS)

1997-07-01

The TECDOC is the result of a series of an advisory and consultants meetings held by the IAEA in 1995-1996 in Vienna (March 1995), in Erlangen Germany (December 1995), in Garching, Germany (June 1996) and in Vienna (November 1996). It was prepared with the participation and contributions of experts from Austria, Canada, Finland, France, Germany, the Republic of Korea, Norway, the Russian Federation, the United Kingdom and the United States of America. The publication not only describes advanced control systems for the improvement of nuclear power plant reliability and efficiency, but also provides a road map to guide interested readers to plan and execute an advanced instrumentation and control project. The subjects include: identification of needs and requirements, justification for safety and user acceptance, and the development of an engineering process. The report should be of interest to nuclear power plant staff, I and C system designers and integrators as well as regulators and researchers. Refs, figs, tabs

Advanced control systems to improve nuclear power plant reliability and efficiency

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-07-01

The TECDOC is the result of a series of an advisory and consultants meetings held by the IAEA in 1995-1996 in Vienna (March 1995), in Erlangen Germany (December 1995), in Garching, Germany (June 1996) and in Vienna (November 1996). It was prepared with the participation and contributions of experts from Austria, Canada, Finland, France, Germany, the Republic of Korea, Norway, the Russian Federation, the United Kingdom and the United States of America. The publication not only describes advanced control systems for the improvement of nuclear power plant reliability and efficiency, but also provides a road map to guide interested readers to plan and execute an advanced instrumentation and control project. The subjects include: identification of needs and requirements, justification for safety and user acceptance, and the development of an engineering process. The report should be of interest to nuclear power plant staff, I and C system designers and integrators as well as regulators and researchers. Refs, figs, tabs.

Evolutionary CANDU 9 plant improvements

International Nuclear Information System (INIS)

Yu, S.K.W.

1999-01-01

The CANDU 9 is a 935 MW(e) nuclear power plant (NPP) based on the multi-unit Darlington and Bruce B designs with additional enhancements from our ongoing engineering and research programs. Added to the advantages of using proven systems and components, CANDU 9 offers improvement features with enhanced safety, improved operability and maintenance including a control centre with advanced man-machine interface, and improved project delivery in both engineering and construction. The CANDU 9 NPP design incorporated safety enhancements through careful attention to emerging licensing and safety issues. The designers assessed, revised and evolved such systems as the moderator, end shield, containment and emergency core cooling (ECC) systems while providing an integrated final design that is more passive and severe-accident-immune. AECL uses a feedback process to incorporate lessons learned from operating plants, from current projects experiences and from the implementation or construction phase of previous projects. Most of the requirements for design improvements are based on a systematic review of current operating CANDU stations in the areas of design and reliability, operability, and maintainability. The CANDU 9 Control Centre provides plant staff with improved operability and maintainability capabilities due to the combination of systematic design with human factors engineering and enhanced operating and diagnostics features. The use of advanced engineering tools and modem construction methods will reduce project implementation risk on project costs and schedules. (author)

Technical feasibility and reliability of passive safety systems for nuclear power plants. Proceedings of an advisory group meeting

International Nuclear Information System (INIS)

1996-12-01

The meeting provided an overview of the key issues on passive safety. Technical problems which may affect future deployment, and the operating experience of passive systems and components, as well as, definitions of passive safety terms, were discussed. Advantages and disadvantages of passive systems were also highlighted. The philosophy behind different passive safety systems was presented and the range of possibility between fully passive and fully active systems was discussed. Refs, figs, tabs

Technical feasibility and reliability of passive safety systems for nuclear power plants. Proceedings of an advisory group meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-01

The meeting provided an overview of the key issues on passive safety. Technical problems which may affect future deployment, and the operating experience of passive systems and components, as well as, definitions of passive safety terms, were discussed. Advantages and disadvantages of passive systems were also highlighted. The philosophy behind different passive safety systems was presented and the range of possibility between fully passive and fully active systems was discussed. Refs, figs, tabs.

Active and Passive Diagnostic Signatures of Special Nuclear Materials

Energy Technology Data Exchange (ETDEWEB)

Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-05-26

An overview will be given discussing signatures associated with special nuclear materials acquired using both active and passive diagnostic techniques. Examples of how technology advancements have helped improve diagnostic capabilities to meet the challenges of today’s applications will be discussed.

Microbial Community Activity And Plant Biomass Are Insensitive To Passive Warming In A Semiarid Ecosystem

Science.gov (United States)

Espinosa, N. J.; Fehmi, J. S.; Rasmussen, C.; Gallery, R. E.

2017-12-01

Soil microorganisms drive biogeochemical and nutrient cycling through the production of extracellular enzymes that facilitate organic matter decomposition and the flux of large amounts of carbon dioxide to the atmosphere. Although dryland ecosystems occupy over 40% of land cover and are projected to expand due to climate change, much of our current understanding of these processes comes from mesic temperate ecosystems. Understanding the responses of these globally predominant dryland ecosystems is therefore important yet complicated by co-occurring environmental changes. For example, the widespread and pervasive transition from grass to woody dominated landscapes is changing the hydrology, fire regimes, and carbon storage potential of semiarid ecosystems. In this study, we used a novel passive method of warming to conduct a warming experiment with added plant debris as either woodchip or biochar, to simulate different long-term carbon additions that accompany woody plant encroachment in semiarid ecosystems. The response of heterotrophic respiration, plant biomass, and microbial activity was monitored bi-annually. We hypothesized that the temperature manipulations would have direct and indirect effects on microbial activity. Warmer soils directly reduce the activity of soil extracellular enzymes through denaturation and dehydration of soil pores and indirectly through reducing microbe-available substrates and plant inputs. Overall, reduction in extracellular enzyme activity may reduce decomposition of coarse woody debris and potentially enhance soil carbon storage in semiarid ecosystems. For all seven hydrolytic enzymes examined as well as heterotrophic respiration, there was no consistent or significant response to experimental warming, regardless of seasonal climatic and soil moisture variation. The enzyme results observed here are consistent with the few other experimental results for warming in semiarid ecosystems and indicate that the controls over soil

2nd ASME-JSME international conference on nuclear engineering -- 1993

International Nuclear Information System (INIS)

Peterson, P.F.

1993-01-01

Volume 1 of this conference covers the following: (1) reactor thermal hydraulic fundamentals; (2) two-phase flow technology; (3) thermal hydraulic phenomena and modeling in nuclear power plants; (4) thermal hydraulic phenomena and modeling in advanced plants; (5) plant transient and accident analysis; (6) performance and reliability evaluation of advanced passive systems; (7) severe accident analysis; (8) severe accident phenomena and experimental investigations; (9) severe accident features and considerations in advanced plants; (10) safety criteria and philosophy; (11) update on VVER safety; (12) safety and reliability features of liquid metal reactor designs; (13) experimental programs in support of advanced reactor plant safety; and (14) passive safety features in advanced plants. Separate abstracts were prepared for 93 papers in this volume

An analysis of the development and application of plant protection UAV based on advanced materials

Science.gov (United States)

Huang, Yuan-hui; Wei, Neng; Quan, Zhi-cheng; Huang, Yu-rong

2018-06-01

The development and application of a number of advanced materials plant protection unmanned aerial vehicle (UAV) is an important part of the comprehensive production of agricultural modernization. The paper is taken as an example of Guangxi No. 1 agricultural service aviation science and Technology Co., Ltd. This paper introduces the internal and external environment of the research and development of the plant protection UAV for the advanced materials of the company. The external environment focuses on the role of the plant protection UAV on the development of the agricultural mechanization; the internal environment focuses on the advantages of the UAV in technology research, market promotion and application, which is imperative. Finally, according to the background of the whole industry, we put forward some suggestions for the developing opportunities and challenges faced by plant protection UAV, hoping to proving some ideas for operators, experts and scholars engaged in agricultural industry.

Advanced nuclear plants: Meeting the economic challenge

International Nuclear Information System (INIS)

Redding, J.R.; MacGregor, P.R.

1993-01-01

As the end of the century draws closer, utilities all over the world face a daunting challenge. Forecasts, such as those prepared by General Electric show a dramatic increase in the need for additional electrical generating capacity (>1000 GW for the world from 1996 through 2006). In the case of the United States, such increases are almost one-third of existing capacity. Furthermore, this capacity must be environmentally benign to satisfy mounting concerns over the environmental impact of burning fossil fuels. This paper discusses economic analyses, similar to least-cost option studies, conducted by General Electric to determine the cost levels that new, advanced nuclear plants must meet to compete with other available generating options

Electronic band-gap modified passive silicon optical modulator at telecommunications wavelengths.

Science.gov (United States)

Zhang, Rui; Yu, Haohai; Zhang, Huaijin; Liu, Xiangdong; Lu, Qingming; Wang, Jiyang

2015-11-13

The silicon optical modulator is considered to be the workhorse of a revolution in communications. In recent years, the capabilities of externally driven active silicon optical modulators have dramatically improved. Self-driven passive modulators, especially passive silicon modulators, possess advantages in compactness, integration, low-cost, etc. Constrained by a large indirect band-gap and sensitivity-related loss, the passive silicon optical modulator is scarce and has been not advancing, especially at telecommunications wavelengths. Here, a passive silicon optical modulator is fabricated by introducing an impurity band in the electronic band-gap, and its nonlinear optics and applications in the telecommunications-wavelength lasers are investigated. The saturable absorption properties at the wavelength of 1.55 μm was measured and indicates that the sample is quite sensitive to light intensity and has negligible absorption loss. With a passive silicon modulator, pulsed lasers were constructed at wavelengths at 1.34 and 1.42 μm. It is concluded that the sensitive self-driven passive silicon optical modulator is a viable candidate for photonics applications out to 2.5 μm.

General description of advanced heavy water reactor

International Nuclear Information System (INIS)

Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

1999-01-01

Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

Advanced PWR technology development -Development of advanced PWR system analysis technology-

Energy Technology Data Exchange (ETDEWEB)

Jang, Moon Heui; Hwang, Yung Dong; Kim, Sung Oh; Yoon, Joo Hyun; Jung, Bub Dong; Choi, Chul Jin; Lee, Yung Jin; Song, Jin Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The primary scope of this study is to establish the analysis technology for the advanced reactor designed on the basis of the passive and inherent safety concepts. This study is extended to the application of these technology to the safety analysis of the passive reactor. The study was performed for the small and medium sized reactor and the large sized reactor by focusing on the development of the analysis technology for the passive components. Among the identified concepts the once-through steam generator, the natural circulation of the integral reactor, heat pipe for containment cooling, and hydraulic valve were selected as the high priority items to be developed and the related studies are being performed for these items. For the large sized passive reactor, the study plans to extend the applicability of the best estimate computer code RELAP5/MOD3 which is widely used for the safety analyses of the reactor system. The improvement and supplementation study of the analysis modeling and the methodology is planned to be carried out for these purpose. The newly developed technologies are expected to be applied to the domestic advanced reactor design and analysis and these technologies will play a key role in extending the domestic nuclear base technology and consolidating self-reliance in the essential nuclear technology. 72 figs, 15 tabs, 124 refs. (Author).

New nuclear plant design and licensing process

International Nuclear Information System (INIS)

Luangdilok, W.

1996-01-01

This paper describes latest developments in the nuclear power reactor technology with emphasis on three areas: (1) the US technology of advanced passive light water reactors (AP600 and S BWR), (2) regulatory processes that certify their safety, and (3) current engineering concerns. The goal is to provide and insight of how the government's regulatory agency guarantees public safety by looking into how new passive safety features were designed and tested by vendors and how they were re-evaluated and retested by the US NRC. The paper then discusses the US 1989 nuclear licensing reform (10 CFR Part 52) whose objectives are to promote the standardization of nuclear power plants and provide for the early and definitive resolution of site and design issues before plants are built. The new licensing process avoids the unpredictability nd escalated construction cost under the old licensing process. Finally, the paper summarizes engineering concerns found in current light water reactors that may not go away in the new design. The concerns are related the material and water chemistry technology in dealing with corrosion problems in water-cooled nuclear reactor systems (PWRs and BWRs). These engineering concerns include core shroud cracking (BWRs), jet pump hold-down beam cracking (BWRs), steam generator tube stress corrosion cracking (PWR)

SWR 1000: An Advanced, Medium-Sized Boiling Water Reactor, Ready for Deployment

International Nuclear Information System (INIS)

Brettschuh, Werner

2006-01-01

The latest developments in nuclear power generation technology mainly concern large-capacity plants in the 1550 -1600 MW range, or very small plants (100 - 350 MW). The SWR 1000 boiling water reactor (BWR), by contrast, offers all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation costs, in the medium-capacity range (1000 - 1250 MW). The SWR 1000 is particularly suitable for countries whose power systems are not designed for large-capacity generating facilities. The economic efficiency of this medium-sized plant in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control (I and C) systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies to be deployed in the SWR 1000 core, meanwhile, have been enlarged from a 10 x 10 rod array to a 12 x 12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70

Problems facing the use of passive safety systems

International Nuclear Information System (INIS)

Burgazzi, L.

2012-01-01

This study will analyze the current state of the art in the reliability of passive systems for extensive use in future nuclear power plants. This case study uncovers the insights on the technological issues associated with the reliability of the systems based on thermal-hydraulics, for which, methods are still in developing phase. The paper is organized as follows: at first the current available methodologies are illustrated and compared, the open issues coming out from their analysis are identified. Five open issues have been identified: 1) the assessment of the uncertainties related to passive system performance; 2) the dependencies among parameters in thermo-hydraulics; 3) the integration of the passive systems within an accident sequence in combination with active systems; 4) the development of dynamic event tree to incorporate the evolution upon time of the physical processes; and 5) the comparison between active and passive systems, mainly on a functional viewpoint. For each open issue the state of the art and the outlook is presented; the relative importance of each of them within the evaluation process is presented as well. (authors)

Fundamental studies of passivity and passivity breakdown

International Nuclear Information System (INIS)

Macdonald, D.D.; Urquidi-Macdonald, M.; Song, H.; Biaggio-Rocha, S.; Searson, P.

1991-11-01

This report summarizes the findings of our fundamental research program on passivity and passivity breakdown. During the past three and one half years in this program (including the three year incrementally-funded grant prior to the present grant), we developed and experimentally tested various physical models for the growth and breakdown of passive films on metal surfaces. These models belong to a general class termed ''point defects models'' (PDMs), in which the growth and breakdown of passive films are described in terms of the movement of anion and cation vacancies

The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

International Nuclear Information System (INIS)

Wang Jianyu; Zhang Shenru; Min Yuanyou

1994-01-01

To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

Experimental and analytical studies on the passive residual heat removal system for the advanced integral type reactor

International Nuclear Information System (INIS)

Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki

2004-01-01

An experiment on the thermal-hydraulic characteristics of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART-P, has been performed, and its experimental results have been analyzed using a best-estimated system analysis code, MARS. The experiment is performed to investigate the performance of the passive residual heat removal system using the high temperature and high pressure thermal-hydraulic test facility (VISTA) which simulates the SMART-P. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfer through the PRHRS heat exchanger in the emergency cooldown tank is sufficient enough to enable a natural circulation of the coolant. Analysis on a typical PRHRS test has been carried out using the MARS code. The overall trends of the calculated flow rate, pressure, temperature, and heat transfer rate in the PRHRS are similar to the experimental data. There is good agreement between the experimental data and the calculated one for the fluid temperature in the PRHRS steam line. However, the calculated fluid temperature in the PRHRS condensate line is higher, the calculated coolant outlet temperature is lower, and the heat transfer rate through the PRHRS heat exchanger is lower than the experimental data. It seems that it is due to an insufficient heat transfer modeling in the pool such as the emergency cooldown tank in the MARS calculation. (author)

Simultaneous Waste Heat and Water Recovery from Power Plant Flue Gases for Advanced Energy Systems

Energy Technology Data Exchange (ETDEWEB)

Wang, Dexin [Gas Technology Inst., Des Plaines, IL (United States)

2016-12-31

This final report presents the results of a two-year technology development project carried out by a team of participants sponsored by the Department of Energy (DOE). The objective of this project is to develop a membrane-based technology to recover both water and low grade heat from power plant flue gases. Part of the recovered high-purity water and energy can be used directly to replace plant boiler makeup water as well as improving its efficiency, and the remaining part of the recovered water can be used for Flue Gas Desulfurization (FGD), cooling tower water makeup or other plant uses. This advanced version Transport Membrane Condenser (TMC) with lower capital and operating costs can be applied to existing plants economically and can maximize waste heat and water recovery from future Advanced Energy System flue gases with CO₂ capture in consideration, which will have higher moisture content that favors the TMC to achieve higher efficiency.

A new advanced software platform for nuclear power plant process information systems

International Nuclear Information System (INIS)

Sorsa, A.

1993-01-01

In the late 80s, ABB Stromberg Power Ltd. started the development of a new generation software platform for the power plant Process Information System (PIS). This development resulted in a software platform called Procontrol PMS. Procontrol PMS is a platform for fully distributed systems which provides the following features: distributed data processing, non-stop architecture, low-cost incremental expansion path, open network architecture, high functionality, effective application development environment, and advanced user interface services. A description of the structure of the Procontrol PMS software is given. ABB has received by May 1992 six orders for nuclear power plant PISs based on Procontrol PMS (4 for PWR plants, 2 for BWRs). The first Procontrol PMS based nuclear power plant PIS was commissioned in 1989 at the Loviisa nuclear power plant and has been running with 100% availability since the commissioning. (Z.S.) 2 figs

Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

International Nuclear Information System (INIS)

Wood, RT

2004-01-01

This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

Thermodynamic analysis of the advanced zero emission power plant

Directory of Open Access Journals (Sweden)

Kotowicz Janusz

2016-03-01

Full Text Available The paper presents the structure and parameters of advanced zero emission power plant (AZEP. This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i oxygen separation from the air through the membrane, (ii combustion of the fuel, and (iii heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC through the main heat recovery steam generator (HRSG. Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

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)

Proceedings of 2017 international congress on advances in nuclear power plants (ICAPP2017)

International Nuclear Information System (INIS)

2017-04-01

The International Congress on Advances in Nuclear Power Plants (ICAPP) provides a forum for leaders of the nuclear industry to exchange information, present results from their work, review the state of the industry, and discuss future directions and needs for the deployment of new nuclear power plant systems around the world. ICAPP will gather industry leaders in several invited lectures in plenary sessions. The theme for ICAPP2017 is 'A New Paradigm in Nuclear Power Safety'. Since the Fukushima Daiichi Accident in 2011, various efforts in improving nuclear safety have been initiated not only in Japan but also in other countries. Decontamination of affected soil and steps toward decommissioning Fukushima Daiichi are proceeding steadily, but many issues to be resolved still remain. Further advances in reactor decommissioning technologies are expected in light of the rising number of old nuclear power plants being closed. The congress also provides an excellent opportunity to discuss these topics. This issue is the collection of 345 papers presented at the entitled meeting. All the 345 papers are indexed individually. (J.P.N.)

Advanced diagnostics and predictive maintenance to improve availability and reliability of ENEL plants

Energy Technology Data Exchange (ETDEWEB)

Cenci, V.; Ghironi, M.; Guidi, L.; Lauro, M.; Pestonesi, D. [ENEL (Italy). Generation and Energy Management Division

2007-07-01

This paper reviews the ENEL Generation and Energy Management strategy for diagnostics and predictive maintenance of power plants and provides a comprehensive description of effective applications and systems. Exploiting the most advanced information and communication technologies makes it possible to capture weak and hidden signals and powerful processing can be used to discover forewarning symptoms and identify anomalies both in the process and, above all, inside the devices. The following systems and applications are presented together with results and impact on plant profitability: expert system for the diagnostics of plant main machinery; advanced diagnostics of 'intelligent' fieldbus devices such as on/off valve motor-driven actuators, control-valve positioners and pneumatic actuators, transmitters; control loop and control valve diagnostics in order to investigate valve friction with an estimation of the residual time to failure; multisensorial diagnostics for coal transport and storage systems aimed at preventing firing and structural damages; and wireless sensor networks for the diagnostics of medium and small size components. 4 refs., 14 figs., 1 tab.

ESBWR-An economical passive plant design

International Nuclear Information System (INIS)

Gonzalez Lopez, A.; Rao, A.

1996-01-01

This paper provides an overview of the design features of the European Simplified Boiling Water Reactor (ESBWR) design. The ESBWR is a plant design that builds on the Simplified Boiling Water Reactor (SBWR) design described in Reference 1 and 2. The major objective of the ESBWR programme is to develop a plant design that utilizes the basic simplicity of the SBWR design features to improve overall economics as discussed in Reference 3. The design is being developed by an international team of utilities, designers and researchers, with the objective of applying it to the utility and regulatory requirements of Europe. (Author)

Performance Assessment of Passive Gaseous Provisions (PGAP). Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

2013-07-01

The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000 on the basis of IAEA General Conference resolution GC(44)/RES/21. INPRO helps to ensure the availability of sustainable nuclear energy in the 21st century and seeks to bring together all interested Member States - both technology holders and technology users - to consider joint actions to achieve desired innovations. To contribute to an international consensus on the definition of the reliability of passive systems that involve natural circulation, and on a methodology to assess this reliability, INPRO initiated a collaborative project on Performance Assessment of Passive Gaseous Provisions (PGAP) in 2007. 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 mitigate the consequences of a severe accident should one occur. However, the reliability of passive safety systems is crucial and must be assessed before they are used extensively 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 a priori unknown. The functions of many passive systems are based on thermohydraulic principles, which until recently were considered as not being subject to any kind of failure. Hence, large and consistent efforts are required to quantify the reliability of such systems. Three participants from three INPRO Member States were involved in this collaborative project. Reliability methods for passive systems (RMPS) and assessment of passive system reliability (APSRA) methodologies were used by the participants to assess the performance and reliability of the passive decay heat removal system of the French gas cooled fast reactor design for station blackout and a loss of coolant accident combined with loss of off-site power, respectively. This publication presents the

NPR and ANSI Containment Study Using Passive Cooling Techniques

International Nuclear Information System (INIS)

Shin, J. J.; Iotti, R. C.; Wright, R. F.

1993-01-01

Passive containment cooling study of NPR (New Production Reactor) and ANSI (Advanced Neutron Source) following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally. All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified Contempt code Contempt-Npr. In order to verify proper heat transfer coefficient, temperature, heat flux and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and halt foot (1.7m) diameter outer cylinder. Comparison of Contempt-Npr and WGOTHIC was done for reduced scale Npr. It is concluded that Npr and ANSI containments can be passively cooled with air alone without extended cooling surfaces or passive water spray

An on-line advanced plant simulator (OLAPS)

International Nuclear Information System (INIS)

Samuels, J.W.

1989-01-01

A PC based on-line advanced plant simulator (OLAPS) for high quality simulations of Portland General Electric's Trojan Nuclear Facility is presented. OLAPS is designed to simulate the thermal-hydraulics of the primary system including core, steam generators, pumps, piping and pressurizer. The simulations are based on a five equation model that has two mass equations, two energy equations, two energy equations, and one momentum equation with a drift flux model to provide closure. A regionwise point reactor kinetics model is used to model the neutron kinetics in the core. The conservation equations, constitutive models and the numerical methods used to solve them are described. OLAPS results are compared with data from chapter 15 of the Trojan Nuclear Facility's final safety analysis report

Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor

International Nuclear Information System (INIS)

Bijlani, C.; Patti, F.; Prasad, V.

1993-01-01

At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions

Advanced nuclear plant design options to cope with external events

International Nuclear Information System (INIS)

2006-02-01

With the stagnation period of nuclear power apparently coming to an end, there is a renewed interest in many Member States in the development and application of nuclear power plants (NPPs) with advanced reactors. Decisions on the construction of several NPPs with evolutionary light water reactors have been made (e.g. EPR Finland for Finland and France) and more are under consideration. There is a noticeable progress in the development and demonstration of innovative high temperature gas cooled reactors, for example, in China, South Africa and Japan. The Generation IV International Forum has defined the International Near Term Deployment programme and, for a more distant perspective, six innovative nuclear energy systems have been selected and certain R and D started by several participating countries. National efforts on design and technology development for NPPs with advanced reactors, both evolutionary and innovative, are ongoing in many Member States. Advanced NPPs have an opportunity to be built at many sites around the world, with very broad siting conditions. There are special concerns that safety of these advanced reactors may be challenged by external events following new scenarios and failure modes, different from those well known for the currently operated reactors. Therefore, the engineering community identified the need to assess the proposed design configurations in relation to external scenarios at the earliest stages of the design development. It appears that an early design optimization in relation to external events is a necessary requirement to achieve safe and economical advanced nuclear power plants. Reflecting on these developments, the IAEA has planned the preparation of a report to define design options for protection from external event impacts in NPPs with evolutionary and innovative reactors. The objective of this publication is to present the state-of-the-art in design approaches for the protection of NPPs with evolutionary and innovative

Trends in advanced reactor development and the role of the IAEA

International Nuclear Information System (INIS)

Semenov, B.; Dastidar, P.; Kupitz, J.; Cleveland, J.; Goodjohn, A.

1992-01-01

This report discusses advanced reactors are being developed for all principal reactor types, i.e. the light and heavy water-cooled reactors, the liquid-metal-cooled reactors and the gas-cooled reactors. Some of these developments are primarily of an evolutionary nature, i.e. they represent improvements in component and system technology, and in construction and operating practices as a result of experience gained with presently operating plants. Other developments are also evolutionary but with some incorporation of innovative features such as providing passive systems for assuring continuous cooling for removal of decay heat from the reactor core. If there is a revival of nuclear power, which may be dictated by ecological and economical factors, advanced reactors now being developed could help to meet the large demand for new plants in developed and developing countries, not only for electricity generation, but also for district heating, desalination and for process heat. The IAEA, as the only global international governmental organization dealing with nuclear power, has promoted international information exchange and international cooperation between all countries with their own advanced nuclear power programmes and has offered assistance to countries with an interest in exploratory or research programmes. In the future the IAEA could play an even more-important role

Issues in risk analysis of passive LWR designs

International Nuclear Information System (INIS)

Youngblood, R.W.; Pratt, W.T.; Amico, P.J.; Gallagher, D.

1992-01-01

This paper discusses issues which bear on the question of how safety is to be demonstrated for ''simplified passive'' light water reactor (LWR) designs. First, a very simplified comparison is made between certain systems in today's plants. comparable systems in evolutionary designs, and comparable systems in the simplified passives. in order to introduce the issues. This discussion is not intended to describe the designs comprehensively, but is offered only to show why certain issues seem to be important in these particular designs. Next, an important class of accident sequences is described; finally, based on this discussion, some priorities in risk analysis are presented and discussed

Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

International Nuclear Information System (INIS)

Ko, A Reum; Lee, Jeong Ik; Yoon, Ho Joon

2016-01-01

Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases

Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

Energy Technology Data Exchange (ETDEWEB)

Ko, A Reum; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-05-15

Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases.

Assessment of fiber optic sensors and other advanced sensing technologies for nuclear power plants

International Nuclear Information System (INIS)

Hashemian, H.M.

1996-01-01

As a result of problems such as calibration drift in nuclear plant pressure sensors and the recent oil loss syndrome in some models of Rosemount pressure transmitters, the nuclear industry has become interested in fiber optic pressure sensors. Fiber optic sensing technologies have been considered for the development of advanced instrumentation and control (I ampersand C) systems for the next generation of reactors and in older plants which are retrofitted with new I ampersand C systems. This paper presents the results of a six-month Phase I study to establish the state-of-the-art in fiber optic pressure sensing. This study involved a literature review, contact with experts in the field, an industrial survey, a site visit to a fiber optic sensor manufacturer, and laboratory testing of a fiber optic pressure sensor. The laboratory work involved both static and dynamic performance tests. This initial Phase I study has recently been granted a two-year extension by the U.S. Nuclear Regulatory Commission (NRC). The next phase will evaluate fiber optic pressure sensors in specific nuclear plant applications in addition to other advanced methods for monitoring critical nuclear plant equipment

Passive solar technology

Energy Technology Data Exchange (ETDEWEB)

Watson, D

1981-04-01

The present status of passive solar technology is summarized, including passive solar heating, cooling and daylighting. The key roles of the passive solar system designer and of innovation in the building industry are described. After definitions of passive design and a summary of passive design principles are given, performance and costs of passive solar technology are discussed. Passive energy design concepts or methods are then considered in the context of the overall process by which building decisions are made to achieve the integration of new techniques into conventional design. (LEW).

Development and utilization of the NRC policy statement on the regulation of advanced nuclear power plants

International Nuclear Information System (INIS)

Williams, P.M.; King, T.L.

1988-06-01

On March 26, 1985, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Policy for Regulation of Advanced Nuclear Power Plants'' (50 FR 11884). This report presents and discusses the Commission's final version of that policy as titled and published on July 8, 1986 ''Regulation of Advanced Nuclear Power Plants, Statement of Policy'' (51 FR 24643). It provides an overview of comments received from the public, of the significant changes from the proposed Policy Statement to the final Policy Statement, and of the Commission's response to six questions contained in the proposed Policy Statement. The report also discusses the definition for advanced reactors, the establishment of an Advanced Reactors Group, the staff review approach and information needs, and the utilization of the Policy Statement in relation to other NRC programs, including the policies for safety goals, severe accidents and standardization. In addition, guidance for advanced reactors with respect to operating experience, technology development, foreign information and data, and prototype testing is provided. Finally, a discussion on the use of less prescriptive and nonprescriptive design criteria for advanced reactors, which the Policy Statement encourages, is presented

Development and validation process of the advanced main control board for next Japanese PWR plants

International Nuclear Information System (INIS)

Tani, M.; Ito, K.; Yokoyama, M.; Imase, M.; Okamoto, H.

2000-01-01

The purpose of main control room improvement is to reduce operator workload and potential human errors by offering a better working environment where operators can maximize their abilities. Japanese pressurized water reactor (PWR) utilities and Mitsubishi group have developed a touch -screen-based main control console (i.e. advanced main control room) the next generation PWRs to further improve the plant operability using a state of the art electronics technology. The advanced main control room consists of an operator console, a supervisor console and large display panels. The functional specifications were evaluated by utility operators using a prototype main control console connected to a plant simulator. (author)

Advanced I and C systems for nuclear power plants feedback of experience

International Nuclear Information System (INIS)

Prehler Heinz Josef

2001-01-01

Advanced I and C systems for nuclear power plants have to meet increasing demands for safety and availability. Additionally specific requirements arising from nuclear qualification have to be fulfilled. To meet both subjects adequately in the future, Siemens has developed advanced I and C technology consisting of the two complementary I and C systems TELEPERM XP and TELEPERM XS. TELEPERM XP is primarily oriented to automation of the non safety related part of the power plant process. Such applications involve extensive open and closed loop control systems and encompass all tasks required for process control via the man-machine interface. Therefore the TELEPERM XP system consists of the AS 620 automation system, the OM 690 process control and management system, the ES 680 engineering system, the DS 670 diagnostic system and the SIMATIC NET bus system. Three versions of automation systems are available: for standard automation, for fail safe automation of safety related tasks and for turbine automation. TELEPERM XS is designed to meet all the requirements on I and C important to safety in nuclear power plants. Typical applications include reactor protection (RPS) and Engineered Safety Features Actuation System functions (ESFAS). TELEPERM XS has been rapidly accepted by the market and has accumulated an extensive operational experience. The expected advantages, namely, reduced space requirements, consistent documentation, improved ergonomics, reduced testing effort, less repair have been confirmed by the operation. The new possibilities to apply intelligent diagnostic methods have been only applied in few cases. Very good service records from a broad field of safety application prove that it is right to use digital I and C systems for safety tasks. The expected advantages such as reduced space requirements, less repairs and less effort for periodic tests, have been confirmed by practical experience. For the future, use of digital I and C systems for safety

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

Passive Strategy with Integrated Passive Safety System (IPSS) for DBAs in SBO

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho; Kim, Jihee; Choi, Jae Young; Jeon, Inseop; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2013-10-15

In this paper, the strategies of coping with DBAs in SBO were proposed by the design with IPSS. Current nuclear power plants adopt emergency strategies using fire truck as a provision of steam generator cooling. However, it has a lot of limitation like water inventory, preparedness and accessibility. In the case of passive strategy by the application of IPSS, faster actions and more efficient performances can be achieved. The application of IPSS implies the preparedness of big water tank which can be used as water supplier, heat sink and filtering medium. The proposed strategies are set under the conservative conditions without AC power. In order to set more realistic and acceptable strategy, the proposed passive strategy has to be combined with the current strategies. The combined strategies can avoid the reiteration and complexity in accidents. Accordingly, the set of operation mode considering action priority with estimating specific conditions is the further work of this research. Removing decay heat is one of the most important issues in safety of nuclear engineering. In the Fukushima accidents, the initial problem was an occurrence of tsunami. It was connected into a station black out (SBO) which lost AC power in site. Finally, SBO with human error induced the failure of decay heat removal. The occurrence of SBO and the failure of decay heat removal imply the questions for solving them. In order to prevent and mitigate SBO, some solutions have been proposed after the Fukushima accident. First of all, physical protection is enhanced to prevent external risks. For example, the tsunami barrier was modified to be higher from 7.5 m to 10 m. The second is to add electrical redundancy to prevent a total loss of electrical power. AAC diesel generators and movable diesel generators are examples for emergency conditions to supply AC power in site. Bunker concept which was proposed in Europe is a representative example. The bunker concept was analyzed to be applied in

Advances in methods for identification and characterization of plant transporter function

DEFF Research Database (Denmark)

Larsen, Bo; Xu, Deyang; Halkier, Barbara Ann

2017-01-01

Transport proteins are crucial for cellular function at all levels. Numerous importers and exporters facilitate transport of a diverse array of metabolites and ions intra- and intercellularly. Identification of transporter function is essential for understanding biological processes at both......-based approaches. In this review, we highlight examples that illustrate how new technology and tools have advanced identification and characterization of plant transporter functions....

NRC review of passive reactor design certification testing programs: Overview, progress, and regulatory perspective

Energy Technology Data Exchange (ETDEWEB)

Levin, A.E.

1995-09-01

New reactor designs, employing passive safety systems, are currently under development by reactor vendors for certification under the U.S. Nuclear Regulatory Commission`s (NRC`s) design certification rule. The vendors have established testing programs to support the certification of the passive designs, to meet regulatory requirements for demonstration of passive safety system performance. The NRC has, therefore, developed a process for the review of the vendors` testing programs and for incorporation of the results of those reviews into the safety evaluations for the passive plants. This paper discusses progress in the test program reviews, and also addresses unique regulatory aspects of those reviews.

FFTF Plant transition mission analysis report

International Nuclear Information System (INIS)

Lund, D.P.

1995-01-01

FFTF (Fast Flux Test Facility) is a 400-MW(t) sodium-cooled, fast flux test reactor at Hanford, designed to test fuels and materials for advanced nuclear power plants; it has no capability for generating electric power. Since a long-term mission could not be found for FFTF, it was placed in standby, and a recommendation was made that it be shut down. Purpose of the FFTF Transition Project is to prepare it for Decontamination and Decommissioning; this will be accomplished by establishing a passively safe and environmentally secure configuration, that can be preserved for several decades. This report presents the results of the mission analysis, which is required by Hanford systems engineering procedures

Advanced Purex process for the new French reprocessing plants

International Nuclear Information System (INIS)

Viala, M.; Ledermann, P.; Pradel, P.

1993-01-01

The paper describes the main process innovations of the new Cogema reprocessing plants of La Hague (UP3 and UP2 800). Major improvements of process like the use of rotary dissolvers and annular columns, and also entirely new processes like solvent distillation and plutonium oxidizing dissolution, yield an advanced Purex process. The results of these innovations are significant improvements for throughput, end-products purification performances and waste minimization. They contribute also to limit personnel exposure. The main results of the first three years of operation are described. (author). 3 refs., 5 figs

The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

International Nuclear Information System (INIS)

Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

1992-07-01

This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

Thermal limits for passive safety of fusion reactors

International Nuclear Information System (INIS)

Kazimi, M.S.; Massidda, J.E.; Oshima, M.

1989-01-01

The thermal response of the first wall and blanket due to power/cooling mismatch in the absence of operation action is examined. The analyses of coolant and power transients are carried out on six reference blanket designs representing a broad range of fusion first wall and blanket technology. It is concluded that the requirement of plant protection will impose sufficiently stringent peak neutron wall loading limits to avoid a serious threat to the public. It is found that for the D-T design,s the operating wall loading may have to be limited to 3 - 8 MW/m/sup 2/ for passive plant protection, depending on the plant design

Interpretation of risk significance of passive component aging using probabilistic structural analysis

International Nuclear Information System (INIS)

Phillips, J.H.; Atwood, C.L.

1993-01-01

The probabilistic risk assessments (PRAs) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. Except as initiating events, the possible failure of passive components is given little consideration. The NRC is sponsoring a project at INEL to investigate the risk significance of passive components as they age. For this project, we developed a technique to calculate the failure probability of passive components over time, and demonstrated the technique by applying it to a weld in the auxiliary feedwater (AFW) system. A decreasing yearly rupture rate for this weld was calculated instead of the increasing rupture rate trend one might expect. We attribute this result to infant mortality; that is, most of those initial flaws that will eventually lead to rupture will do so early in life. This means that although each weld in a population may be wearing out, the population as a whole can exhibit a decreasing rupture rate. This observation has implications for passive components in commercial nuclear plants and other facilities where aging is a concern. For the population of passive components that exhibit a decreasing failure rate, risk increase is not a concern. The next step of the work is to identify the attributes that contribute to this decreasing rate, and to determine any attributes that would contribute to an increasing failure rate and thus to an increased risk

Advanced Neutron Source: Plant Design Requirements

International Nuclear Information System (INIS)

1990-07-01

The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS

Passive safety systems for integral reactors

International Nuclear Information System (INIS)

Kuul, V.S.; Samoilov, O.B.

1996-01-01

In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs

Passive safety systems for integral reactors

Energy Technology Data Exchange (ETDEWEB)

Kuul, V S; Samoilov, O B [OKB Mechanical Engineering (Russian Federation)

1996-12-01

In this paper, a wide range of passive safety systems intended for use on integral reactors is considered. The operation of these systems relies on natural processes and does not require external power supplies. Using these systems, there is the possibility of preventing serious consequences for all classes of accidents including reactivity, loss-of-coolant and loss of heat sink as well as severe accidents. Enhancement of safety system reliability has been achieved through the use of self-actuating devices, capable of providing passive initiation of protective and isolation systems, which respond immediately to variations in the physical parameters of the fluid in the reactor or in a guard vessel. For beyond design base accidents accompanied by complete loss of heat removal capability, autonomous self-actuated ERHR trains have been proposed. These trains are completely independent of the secondary loops and need no action to isolate them from the steam turbine plant. Passive safety principles have been consistently implemented in AST-500, ATETS-200 and VPBER 600 which are new generation NPPs developed by OKBM. Their main characteristic is enhanced stability over a wide range of internal and external emergency initiators. (author). 10 figs.

Issues and approaches in risk-based aging analyses of passive components

International Nuclear Information System (INIS)

Uryasev, S.P.; Samanta, P.K.; Vesely, W.E.

1994-01-01

In previous NRC-sponsored work a general methodology was developed to quantify the risk contributions from aging components at nuclear plants. The methodology allowed Probabilistic Risk Analyses (PRAs) to be modified to incorporate the age-dependent component failure rates and also aging maintenance models to evaluate and prioritize the aging contributions from active components using the linear aging failure rate model and empirical components aging rates. In the present paper, this methodology is extended to passive components (for example, the pipes, heat exchangers, and the vessel). The analyses of passive components bring in issues different from active components. Here, we specifically focus on three aspects that need to be addressed in risk-based aging prioritization of passive components

Operating boundaries of full-scale advanced water reuse treatment plants: many lessons learned from pilot plant experience.

Science.gov (United States)

Bele, C; Kumar, Y; Walker, T; Poussade, Y; Zavlanos, V

2010-01-01

Three Advanced Water Treatment Plants (AWTP) have recently been built in South East Queensland as part of the Western Corridor Recycled Water Project (WCRWP) producing Purified Recycled Water from secondary treated waste water for the purpose of indirect potable reuse. At Luggage Point, a demonstration plant was primarily operated by the design team for design verification. The investigation program was then extended so that the operating team could investigate possible process optimisation, and operation flexibility. Extending the demonstration plant investigation program enabled monitoring of the long term performance of the microfiltration and reverse osmosis membranes, which did not appear to foul even after more than a year of operation. The investigation primarily identified several ways to optimise the process. It highlighted areas of risk for treated water quality, such as total nitrogen. Ample and rapid swings of salinity from 850 to 3,000 mg/l-TDS were predicted to affect the RO process day-to-day operation and monitoring. Most of the setpoints used for monitoring under HACCP were determined during the pilot plant trials.

Impurity diffusion, point defect engineering, and surface/interface passivation in germanium

KAUST Repository

Chroneos, Alexander I.; Schwingenschlö gl, Udo; Dimoulas, Athanasios Dimoulas

2012-01-01

in view of recent results. The importance of electrically active defects on the Ge surface and interfaces is addressed considering strategies to suppress them and to passivate the surfaces/interfaces, bearing in mind their importance for advanced devices

Advancing the Use of Passive Sampling in Risk Assessment and Management of Sediments Contaminated with Hydrophobic Organic Chemicals: Results of an International Ex Situ Passive Sampling Interlaboratory Comparison

Science.gov (United States)

This work presents the results of an international interlaboratory comparison on ex situ passive sampling in sediments. The main objectives were to map the state of the science in passively sampling sediments, identify sources of variability, provide recommendations and practica...

Hybrid passivated colloidal quantum dot solids

KAUST Repository

Ip, Alex

2012-07-29

Colloidal quantum dot (CQD) films allow large-area solution processing and bandgap tuning through the quantum size effect. However, the high ratio of surface area to volume makes CQD films prone to high trap state densities if surfaces are imperfectly passivated, promoting recombination of charge carriers that is detrimental to device performance. Recent advances have replaced the long insulating ligands that enable colloidal stability following synthesis with shorter organic linkers or halide anions, leading to improved passivation and higher packing densities. Although this substitution has been performed using solid-state ligand exchange, a solution-based approach is preferable because it enables increased control over the balance of charges on the surface of the quantum dot, which is essential for eliminating midgap trap states. Furthermore, the solution-based approach leverages recent progress in metal:chalcogen chemistry in the liquid phase. Here, we quantify the density of midgap trap states in CQD solids and show that the performance of CQD-based photovoltaics is now limited by electrong-"hole recombination due to these states. Next, using density functional theory and optoelectronic device modelling, we show that to improve this performance it is essential to bind a suitable ligand to each potential trap site on the surface of the quantum dot. We then develop a robust hybrid passivation scheme that involves introducing halide anions during the end stages of the synthesis process, which can passivate trap sites that are inaccessible to much larger organic ligands. An organic crosslinking strategy is then used to form the film. Finally, we use our hybrid passivated CQD solid to fabricate a solar cell with a certified efficiency of 7.0%, which is a record for a CQD photovoltaic device. © 2012 Macmillan Publishers Limited. All rights reserved.

Hybrid passivated colloidal quantum dot solids

KAUST Repository

Ip, Alex; Thon, Susanna; Hoogland, Sjoerd H.; Voznyy, Oleksandr; Zhitomirsky, David; Debnath, Ratan K.; Levina, Larissa; Rollny, Lisa R.; Carey, Graham H.; Fischer, Armin H.; Kemp, Kyle W.; Kramer, Illan J.; Ning, Zhijun; Labelle, André J.; Chou, Kang Wei; Amassian, Aram; Sargent, E. H.

2012-01-01

Colloidal quantum dot (CQD) films allow large-area solution processing and bandgap tuning through the quantum size effect. However, the high ratio of surface area to volume makes CQD films prone to high trap state densities if surfaces are imperfectly passivated, promoting recombination of charge carriers that is detrimental to device performance. Recent advances have replaced the long insulating ligands that enable colloidal stability following synthesis with shorter organic linkers or halide anions, leading to improved passivation and higher packing densities. Although this substitution has been performed using solid-state ligand exchange, a solution-based approach is preferable because it enables increased control over the balance of charges on the surface of the quantum dot, which is essential for eliminating midgap trap states. Furthermore, the solution-based approach leverages recent progress in metal:chalcogen chemistry in the liquid phase. Here, we quantify the density of midgap trap states in CQD solids and show that the performance of CQD-based photovoltaics is now limited by electrong-"hole recombination due to these states. Next, using density functional theory and optoelectronic device modelling, we show that to improve this performance it is essential to bind a suitable ligand to each potential trap site on the surface of the quantum dot. We then develop a robust hybrid passivation scheme that involves introducing halide anions during the end stages of the synthesis process, which can passivate trap sites that are inaccessible to much larger organic ligands. An organic crosslinking strategy is then used to form the film. Finally, we use our hybrid passivated CQD solid to fabricate a solar cell with a certified efficiency of 7.0%, which is a record for a CQD photovoltaic device. © 2012 Macmillan Publishers Limited. All rights reserved.

Proteomics and plant disease: advances in combating a major threat to the global food supply.

Science.gov (United States)

Rampitsch, Christof; Bykova, Natalia V

2012-02-01

The study of plant disease and immunity is benefiting tremendously from proteomics. Parallel streams of research from model systems, from pathogens in vitro and from the relevant pathogen-crop interactions themselves have begun to reveal a model of how plants succumb to invading pathogens and how they defend themselves without the benefit of a circulating immune system. In this review, we discuss the contribution of proteomics to these advances, drawing mainly on examples from crop-fungus interactions, from Arabidopsis-bacteria interactions, from elicitor-based model systems and from pathogen studies, to highlight also the important contribution of non-crop systems to advancing crop protection. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Development of Advanced Concept for Shortening Construction Period of ABWR Plant

International Nuclear Information System (INIS)

Hiroshi Ijichi; Toshio Yamashita; Masahiro Tsutagawa; Hiroya Mori; Nobuaki Ooshima; Jun Miura; Minoru Kanechika; Nobuaki Miura

2002-01-01

Construction of a nuclear power plant (NPP) requires a very long period because of large amount of construction materials and many issues for negotiation among multiple sections. Shortening the construction period advances the date of return on an investment, and can also result in reduced construction cost. Therefore, the study of this subject has a very high priority for utilities. We achieved a construction period of 37 months from the first concrete work to fuel loading (F/L) (51.5 months from the inspection of the foundation (I/F) to the start of commercial operation (C/O)) at the Kashiwazaki-Kariwa NPPs No. 6 and 7 (KK-6/7), which are the first ABWR plants in the world. At TEPCO's next plant, we think that a construction period of less than 36 months (45 months from I/F to C/O) can be realized based on conventional methods such as early start of equipment installation and blocking of equipment to be brought in advance. Furthermore, we are studying the feasibility of a 21.5-month construction period (30 months from I/F to C/O) with advanced ideas and methods. The important concepts for a 21.5-month construction period are adoption of a new building structure that is the steel plate reinforced concrete (SC) structure and promotion of extensive modularization of equipment and building structure. With introducing these new concepts, we are planning the master schedule (M/S) and finding solutions to conflicts in the schedule of area release from building construction work to equipment installation work (schedule-conflicts.) In this report, we present the shortest construction period and an effective method to put it into practice for the conventional general arrangement (GA) of ABWR. In the future, we will continue the study on the improvement of building configuration and arrangements, and make clear of the concept for large composite modules of building structures and equipment. (authors)

Application of Advanced Technology to Improve Plant Performance in Nuclear Power Plants

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

Advanced Active Acoustics Lab (AAAL)

Data.gov (United States)

Federal Laboratory Consortium — The Advanced Active Acoustics Lab (AAAL) is a state-of-the-art Undersea Warfare (USW) acoustic data analysis facility capable of both active and passive underwater...

Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

International Nuclear Information System (INIS)

Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

1988-01-01

A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance

Experimental and analytical studies of a passive shutdown heat removal system for advanced LMRs

Energy Technology Data Exchange (ETDEWEB)

Heineman, J.; Kraimer, M.; Lottes, P.; Pedersen, D.; Stewart, R.; Tessier, J.

1988-01-01

A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance.

Advanced Grid-Friendly Controls Demonstration Project for Utility-Scale PV Power Plants

Energy Technology Data Exchange (ETDEWEB)

Gevorgian, Vahan; O' Neill, Barbara

2016-01-21

A typical photovoltaic (PV) power plant consists of multiple power electronic inverters and can contribute to grid stability and reliability through sophisticated 'grid-friendly' controls. The availability and dissemination of actual test data showing the viability of advanced utility-scale PV controls among all industry stakeholders can leverage PV's value from being simply an energy resource to providing additional ancillary services that range from variability smoothing and frequency regulation to power quality. Strategically partnering with a selected utility and/or PV power plant operator is a key condition for a successful demonstration project. The U.S. Department of Energy's (DOE's) Solar Energy Technologies Office selected the National Renewable Energy Laboratory (NREL) to be a principal investigator in a two-year project with goals to (1) identify a potential partner(s), (2) develop a detailed scope of work and test plan for a field project to demonstrate the gird-friendly capabilities of utility-scale PV power plants, (3) facilitate conducting actual demonstration tests, and (4) disseminate test results among industry stakeholders via a joint NREL/DOE publication and participation in relevant technical conferences. The project implementation took place in FY 2014 and FY 2015. In FY14, NREL established collaborations with AES and First Solar Electric, LLC, to conduct demonstration testing on their utility-scale PV power plants in Puerto Rico and Texas, respectively, and developed test plans for each partner. Both Puerto Rico Electric Power Authority and the Electric Reliability Council of Texas expressed interest in this project because of the importance of such advanced controls for the reliable operation of their power systems under high penetration levels of variable renewable generation. During FY15, testing was completed on both plants, and a large amount of test data was produced and analyzed that demonstrates the ability of

Distributed Control Systems in New Nuclear Power Plants

International Nuclear Information System (INIS)

Doerfler, Joseph

2008-01-01

With the growing demand for energy many countries have expressed interest in constructing new plants over the next 15 to 20 years. These expectations have presented a challenge to the nuclear industry to provide a high volume of construction. A key strategy to meet this challenge is developing an advanced nuclear power plant design that allows for a modular construction, a high level of standardization, passive safety features, reduced number of components, and a short bid-to-build time. In addition, the implementation of the plant control system has evolved as new technologies emerge to support these goals. The purpose of this paper is to discuss the ways that the distributed control and information systems in the new generation of nuclear power plants will differ from those currently in service. The new designs provide opportunities to improve overall performance through the use of bus technology, a video display driven Human System Interface, enhanced diagnostics and improved maintenance features. However, the new technologies must fully address requirements for cyber security and high reliability. This paper will give an overview of new technology, improvements, as well as emerging issues in new plant design. (authors)

Distributed Control Systems in New Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Doerfler, Joseph [Westinghouse Electric Company, 4350 Northern Pike, Monroeville, PA 15146 (United States)

2008-07-01

With the growing demand for energy many countries have expressed interest in constructing new plants over the next 15 to 20 years. These expectations have presented a challenge to the nuclear industry to provide a high volume of construction. A key strategy to meet this challenge is developing an advanced nuclear power plant design that allows for a modular construction, a high level of standardization, passive safety features, reduced number of components, and a short bid-to-build time. In addition, the implementation of the plant control system has evolved as new technologies emerge to support these goals. The purpose of this paper is to discuss the ways that the distributed control and information systems in the new generation of nuclear power plants will differ from those currently in service. The new designs provide opportunities to improve overall performance through the use of bus technology, a video display driven Human System Interface, enhanced diagnostics and improved maintenance features. However, the new technologies must fully address requirements for cyber security and high reliability. This paper will give an overview of new technology, improvements, as well as emerging issues in new plant design. (authors)

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

Design related aspects in advanced nuclear fission plants

International Nuclear Information System (INIS)

Hoffelner, Wolfgang

2011-01-01

Important issues to be considered for design of future reactors are: extrapolation of stress rupture data, creep-fatigue, negligible creep, damage monitoring. The paper highlights some new developments taking examples from a martensitic steel (mod 9% Cr), oxide dispersion strengthened (ODS) steels and nickel-base superalloys. Traditional approaches to extrapolation of (thermal) stress rupture data like Larson-Miller Parameter or Monkman-Grant rule seem to be valid concepts also for advanced reactors. However, a significant influence of cyclic softening on creep rates and stress rupture data can be expected as shown for grade 91. This is particularly true for creep-fatigue interactions. Based on cyclic stress-strain behaviour it is also possible to get very good life-time predictions under creep-fatigue with a strain range separation (inelastic fatigue and creep ranges) technique which could replace the currently used linear life fraction rule. Results from in-beam irradiation creep reveal no significant influence of dispersoid size. It can be assumed that irradiation creep is a matrix property. Finally it is shown that micro-sample testing of exposed material could be used as an advanced method for damage assessment in future nuclear power plants.

Nature of bonding forces between two hydrogen-passivated silicon wafers

DEFF Research Database (Denmark)

Stokbro, Kurt; Nielsen, E.; Hult, E.

1998-01-01

The nature and strength of the bonding forces between two II-passivated Si surfaces are studied with the density-functional theory, using an approach based on recent theoretical advances in understanding of van der Waals forces between two surfaces. Contrary to previous suggestions of van der Waals...

Overview of the US program of controls for advanced reactors

International Nuclear Information System (INIS)

White, J.D.; Sackett, J.I.; Monson, R.; Lindsay, R.W.; Carroll, D.G.

1989-01-01

An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed

Passive cooling system for nuclear reactor containment structure

Science.gov (United States)

Gou, Perng-Fei; Wade, Gentry E.

1989-01-01

A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Development of Passive Fuel Cell Thermal Management Heat Exchanger

Science.gov (United States)

Burke, Kenneth A.; Jakupca, Ian J.; Colozza, Anthony J.

2010-01-01

The NASA Glenn Research Center is developing advanced passive thermal management technology to reduce the mass and improve the reliability of space fuel cell systems for the NASA Exploration program. The passive thermal management system relies on heat conduction within highly thermally conductive cooling plates to move the heat from the central portion of the cell stack out to the edges of the fuel cell stack. Using the passive approach eliminates the need for a coolant pump and other cooling loop components within the fuel cell system which reduces mass and improves overall system reliability. Previous development demonstrated the performance of suitable highly thermally conductive cooling plates that could conduct the heat, provide a sufficiently uniform temperature heat sink for each cell of the fuel cell stack, and be substantially lighter than the conventional thermal management approach. Tests were run with different materials to evaluate the design approach to a heat exchanger that could interface with the edges of the passive cooling plates. Measurements were made during fuel cell operation to determine the temperature of individual cooling plates and also to determine the temperature uniformity from one cooling plate to another.

A technique of including the effect of aging of passive components in probabilistic risk assessments

International Nuclear Information System (INIS)

Phillips, J.H.; Weidenhamer, G.H.

1992-01-01

The probabilistic risk assessments (PRAS) being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing methods for selecting risk-significant passive components and including them in PRAS. These methods provide effective ways to prioritize passive components for inspection, and where inspection reveals aging damage, mitigation or repair can be employed to reduce the likelihood of component failure. We demonstrated a method by selecting a weld in the auxiliary feedwater (AFW) system, basing our selection on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then modified and used the Piping Reliability Analysis Including Seismic Events (PRAISE) computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The PRAISE code was modified to include the effects of changing design material properties with age and changing stress cycles. The calculation included the effects of mechanical loads and thermal transients typical of the service loads for this piping design and the effects of thermal cycling caused by a leaking check valve. However, this particular calculation showed little change in low component failure probability and plant risk for 48 years of service. However, sensitivity studies showed that if the probability of component failure is high, the effect on plant risk is significant. The success of this demonstration shows that this method could be applied to nuclear power plants. The demonstration showed the method is too involved (PRAISE takes a long time to perform the calculation and the input information is extensive) for handling a large number of passive components and therefore simpler methods are needed