WorldWideScience

Sample records for passive safety systems

  1. Considerations on nuclear reactor passive safety systems

    International Nuclear Information System (INIS)

    2016-01-01

    After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

  2. 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

  3. 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.

  4. Preliminary safety evaluation for CSR1000 with passive safety system

    International Nuclear Information System (INIS)

    Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

    2014-01-01

    Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities

  5. 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)

  6. 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

  7. 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

  8. Reliability of thermal-hydraulic passive safety systems

    International Nuclear Information System (INIS)

    D'Auria, F.; Araneo, D.; Pierro, F.; Galassi, G.

    2014-01-01

    The scholar will be informed of reliability concepts applied to passive system adopted for nuclear reactors. Namely, for classical components and systems the failure concept is associated with malfunction of breaking of hardware. In the case of passive systems the failure is associated with phenomena. A method for studying the reliability of passive systems is discussed and is applied. The paper deals with the description of the REPAS (Reliability Evaluation of Passive Safety System) methodology developed by University of Pisa (UNIPI) and with results from its application. The general objective of the REPAS methodology is to characterize the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems

  9. TEPSS - Technology Enhancement for Passive Safety Systems

    International Nuclear Information System (INIS)

    Hart, J.; Slegers, W.J.M.; Boer, S.L. de; Huggenberger, M.; Lopez Jimenez, J.; Munoz-Cabo Gonzalez, J.L.; Reventos Puigjaner, F.

    2000-01-01

    The objective of the TEPSS project was to make significant additions to the technology base of the European Simplified Boiling Water Reactor (ESBWR). The project focused on mixing and stratification phenomena in large water pools, passive decay heat removal from containments, and effects of aerosol deposition inside a passive heat exchanger. The PSI experimental facility LINX (Large-scale Investigation of Natural Circulation and Mixing) has been used to investigate venting of steam and steam-noncondensable gas mixtures into water pools. The test revealed that no significant steam bypass could be detected when injecting a mixture of steam or air and that mixing was very efficient. In addition to the tests, 3-D numerical computations and initial model development have been performed to study the behaviour of bubble plumes in water pools. The major part of the TEPSS project studied selective aspects of the response technology of modem pressure-suppression type containment designs and of passive-type decay heat removal systems. The work included an experimental phase using the large-scale experimental facility PANDA (Passive Nachwaermeabfuhr und Druckabbau), operated by PSI, where eight experiments successfully have been executed to test the performance of the ESBWR containment configuration. The PANDA tests have been analysed successfully using thermalhydraulic system analysis codes and 3-D CFD codes. The AIDA (Aerosol Impaction and Deposition Analysis) experimental facility of PSI has been used to investigate the degradation of passive decay heat removal due to fission product aerosols deposited on the inside surfaces of the PCC (Passive Containment Cooler) heat exchanger tubes. The one test performed revealed that the degradation of the heat transfer in the PCC tubes due to the deposition of aerosols reached about 20%. The test has been analysed using the MELCOR severe accident analysis code. (author)

  10. Passive components of NPP safety-related systems

    International Nuclear Information System (INIS)

    Ionaytis Romuald, R.; Bubnova Tatyana, A.

    2005-01-01

    This paper presents a new passive components with having drives: fast-response cutoff valves; modular actuators with opposite cocking pneumatic drives and actuation spring drives; voting electromagnetic valve units for control of pneumatic drives; passive initiators of actuation; visual diagnostics . All these devices have been developed and tested at mock-ups. This paper presents also the following direct-action passive safety components: modular pressure-relief safety valves; pilot safety valves with passive action; check valves with remote position indicator and after-tightening; modular inserts for limiting emergency coolant flow; vortex rectifier; critical weld fasteners; gas-liquid valves; fast-removable seal assembly; seal spring loaders; grooves for increasing hydraulic resistance. Replacement of active safety system components for passive ones improves the general reliability NPP by 1.5 or 2 orders of magnitudes. (authors)

  11. 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

  12. 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

  13. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  14. Passive safety; Passive Sicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Rueckert, J. [Skoda Auto a.s., Mlada Boleslav (Czech Republic). Interieurentwicklung und Versuche; Hau, M. [Skoda Auto a.s., Mlada Boleslav (Czech Republic). Koordination der Fahrzeugsicherung

    2004-05-01

    The specifications for passive safety are partly based on the legal requirements for all export markets combined with the strict internal standards of Volkswagen Group. The Euro NCAP tests and their precisely defined testing methods using the new point assessment are very important. (orig.)

  15. 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)

  16. A reliability assessment methodology for the VHTR passive safety system

    International Nuclear Information System (INIS)

    Lee, Hyungsuk; Jae, Moosung

    2014-01-01

    The passive safety system of a VHTR (Very High Temperature Reactor), which has recently attracted worldwide attention, is currently being considered for the design of safety improvements for the next generation of nuclear power plants in Korea. The functionality of the passive system does not rely on an external source of an electrical support system, but on the intelligent use of natural phenomena. Its function involves an ultimate heat sink for a passive secondary auxiliary cooling system, especially during a station blackout such as the case of the Fukushima Daiichi reactor accidents. However, it is not easy to quantitatively evaluate the reliability of passive safety for the purpose of risk analysis, considering the existing active system failure since the classical reliability assessment method cannot be applied. Therefore, we present a new methodology to quantify the reliability based on reliability physics models. This evaluation framework is then applied to of the conceptually designed VHTR in Korea. The Response Surface Method (RSM) is also utilized for evaluating the uncertainty of the maximum temperature of nuclear fuel. The proposed method could contribute to evaluating accident sequence frequency and designing new innovative nuclear systems, such as the reactor cavity cooling system (RCCS) in VHTR to be designed and constructed in Korea.

  17. LOCA analysis of SCWR-M with passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

    2013-06-15

    Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

  18. Reduced scale PWR passive safety system designing by genetic algorithms

    International Nuclear Information System (INIS)

    Cunha, Joao J. da; Alvim, Antonio Carlos M.; Lapa, Celso Marcelo Franklin

    2007-01-01

    This paper presents the concept of 'Design by Genetic Algorithms (DbyGA)', applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement

  19. 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

  20. 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.

  1. 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

  2. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. 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)

  4. Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

  5. 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.

  6. 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.

  7. A Novel Control Algorithm for Integration of Active and Passive Vehicle Safety Systems in Frontal Collisions

    Directory of Open Access Journals (Sweden)

    Daniel Wallner

    2010-10-01

    Full Text Available The present paper investigates an approach to integrate active and passive safety systems of passenger cars. Worldwide, the introduction of Integrated Safety Systems and Advanced Driver Assistance Systems (ADAS is considered to continue the today

  8. System code improvements for modelling passive safety systems and their validation

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Sebastian; Cron, Daniel von der; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    GRS has been developing the system code ATHLET over many years. Because ATHLET, among other codes, is widely used in nuclear licensing and supervisory procedures, it has to represent the current state of science and technology. New reactor concepts such as Generation III+ and IV reactors and SMR are using passive safety systems intensively. The simulation of passive safety systems with the GRS system code ATHLET is still a big challenge, because of non-defined operation points and self-setting operation conditions. Additionally, the driving forces of passive safety systems are smaller and uncertainties of parameters have a larger impact than for active systems. This paper addresses the code validation and qualification work of ATHLET on the example of slightly inclined horizontal heat exchangers, which are e. g. used as emergency condensers (e. g. in the KERENA and the CAREM) or as heat exchanger in the passive auxiliary feed water systems (PAFS) of the APR+.

  9. Expansion of passive safety function

    International Nuclear Information System (INIS)

    Inai, Nobuhiko; Nei, Hiromichi; Kumada, Toshiaki.

    1995-01-01

    Expansion of the use of passive safety functions is proposed. Two notions are presented. One is that, in the design of passive safety nuclear reactors where aversion of active components is stressed, some active components are purposely introduced, by which a system is built in such a way that it behaves in an apparently passive manner. The second notion is that, instead of using a passive safety function alone, a passive safety function is combined with some active components, relating the passivity in the safety function with enhanced controllability in normal operation. The nondormant system which the authors propose is one example of the first notion. This is a system in which a standby safety system is a portion of the normal operation system. An interpretation of the nondormant system via synergetics is made. As an example of the second notion, a PIUS density lock aided with active components is proposed and is discussed

  10. 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

  11. 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.

  12. Modelling of Condensation in Vertical Tubes for Passive Safety System

    International Nuclear Information System (INIS)

    Papini, D.; Ricotti, M.; Santini, L.; Grgic, D.

    2008-01-01

    Condensation in vertical tubes plays an important role in the performance of heat exchangers in passive safety systems, widely adopted in next generation reactors. Vertical pipe condensers are implemented in the GE-SBWR1000 Isolation Condenser as well as in the Emergency Heat Removal System (EHRS) of the IRIS reactor. The transient and safety analysis is usually carried out by means of best-estimate, thermalhydraulic codes, as RELAP. Suitable heat transfer correlations are required to duly model the two-phase processes. As far as the condensation process is concerned, RELAP5/MOD3.3 adopts the Nusselt correlation to calculate the heat transfer coefficient in laminar conditions and the Shah correlation for turbulent conditions; the maximum of the predictions from laminar and turbulent regimes is used to calculate the condensation heat transfer coefficient. Shah correlation is generally considered as the best empirical correlation for turbulent annular film condensation, but suitable in proper ranges of the various parameters. Nevertheless, recent investigations have pointed out that its validity is highly questionable for high pressure and large diameter tube applications with water, as should be for the utilization for vertical tube condensers in passive safety systems. Thus, a best-estimate model, based on the theory of film condensation on a plain wall, is proposed. Condensate velocity, expressed in terms of Reynolds number, governs the development of three different regime zones: laminar, laminar wavy and turbulent. The best correlation for each regime (Nusselt's for laminar, Kutateladze's for laminar wavy and Chen's for turbulent) is considered and then implemented in RELAP code. Comparison between the Nusselt-Shah and the proposed model shows substantial differences in heat transfer coefficient prediction. Especially, a trend of increasing value of the heat transfer coefficient with tube abscissa (and quality decreasing) is predicted, when turbulence

  13. 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

    Beznau nuclear power plant. The third solution is to make a cooling redundancy. Integrated passive safety system (IPSS) proposed in 2011 is a representative design adding big water tanks outside containment on a top of auxiliary building as a heat sink and water supplier. Three methods are each merit and demerit. After the design concept of IPSS was proposed, main functions were analyzed to be verified. With the simulated performances of IPSS in conservative conditions, passive counter strategies to cope with design based accidents (DBAs) were proposed and described in a basic condition of a SBO.

  14. 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)

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. 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

  19. 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

  20. 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)

  1. 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

  2. Key issues for passive safety

    International Nuclear Information System (INIS)

    Hayns, M.R.

    1996-01-01

    The paper represents a summary of the introductory presentation made at this Advisory Group Meeting on the Technical Feasibility and Reliability of Passive Safety Systems. It was intended as an overview of our views on what are the key issues and what are the technical problems which might dominate any future developments of passive safety systems. It is, therefore, not a ''review paper'' as such and only record the highlights. (author)

  3. Key issues for passive safety

    Energy Technology Data Exchange (ETDEWEB)

    Hayns, M R [AEA Technology, Harwell, Didcot (United Kingdom). European Institutions; Hicken, E F [Forschungszentrum Juelich GmbH (Germany)

    1996-12-01

    The paper represents a summary of the introductory presentation made at this Advisory Group Meeting on the Technical Feasibility and Reliability of Passive Safety Systems. It was intended as an overview of our views on what are the key issues and what are the technical problems which might dominate any future developments of passive safety systems. It is, therefore, not a ``review paper`` as such and only record the highlights. (author).

  4. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  5. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  6. 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

  7. 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.

  8. Inherent and passive safety measures in accelerator driven systems: a safety strategy for ADS

    International Nuclear Information System (INIS)

    Maschek, W.; Rineiski, A.; Morita, K.; Flad, M.

    2001-01-01

    The efficiency of Accelerator Driven Systems (ADSs) for the transmutation and incineration of nuclear waste is strongly related to the utilization of so-called dedicated fuels. In the ideal case these fuels should consist of pure TRUs without fertile materials as 238 U or 232 Th to achieve highest incineration/transmutation rates. Dedicated fuels still have to be developed and programs are under way for their fabrication, irradiation and testing. These fertile-free fuels may suffer from deteriorated thermal or thermo-mechanical properties, as a reduced melting point, reduced thermal conductivity or even thermal instability. First analyses have shown that the use of dedicated fuels may lead to a strong deterioration of the safety parameters of the reactor core as e.g. the void worth, the Doppler or the kinetics quantities as neutron generation time and β eff . In addition, a dedicated core may contain multiple ''critical'' fuel masses, resulting in a considerable recriticality potential. Current knowledge on these dedicated fuels suggests that ''critical'' reactors may not be feasible, because of safety reasons. However, for ADSs, the salient hope has been promoted that due to the subcriticality of the system the poor safety features of such fuels could be coped with. Analyses are presented which show potential safety problems for such dedicated cores. Respecting the results of these analyses a safety strategy is proposed along the lines of defense approach in analogy with ideas formerly developed for fast reactors. Inherent and passive safety measures are integrated into the various defense lines. (author)

  9. Natural circulation and stratification in the various passive safety systems of the SWR 1000

    International Nuclear Information System (INIS)

    Meseth, J.

    2002-01-01

    In some of the passive safety systems of Siemens' SWR 1000 boiling water reactor (i.e. the emergency condensers and containment cooling condensers), natural circulation is the main effect on both the primary and secondary sides by which optimum system efficiency is achieved. Other passive safety systems of the SWR 1000 require natural circulation on the secondary side only (condensation of steam discharged by the safety and relief valves; cooling of the Reactor Pressure Vessel (RPV) by flooding from the outside in case of core melt), while still other systems require stratification to be effective (i.e. the passive pressure pulse transmitters and steam-driven scram tanks). Complex natural circulation and stratification can take place simultaneously if fluids with different densities are enclosed in a single volume (in a core melt accident, for example, the nitrogen, steam and hydrogen in the containment). Related problems and the solutions thereto planned for the SWR 1000 are reported from the designer's viewpoint. (author)

  10. Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

    Directory of Open Access Journals (Sweden)

    Franco Pierro

    2009-01-01

    Full Text Available The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks' formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

  11. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    International Nuclear Information System (INIS)

    Kim, Sang Ho; Chang, Soon Heung; Choi, Yu Jung; Jeong, Yong Hoon

    2015-01-01

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  12. A passive decay heat removal strategy of the integrated passive safety system (IPSS) for SBO combined with LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Ho [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of); Chang, Soon Heung [Handong Global University, 558, Handong-ro, Buk-gu, Pohang Gyeongbuk 37554 (Korea, Republic of); Choi, Yu Jung [Korea Hydro and Nuclear Power Co.—Central Research Institute, 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 34101 (Korea, Republic of); Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291, Daehak-ro, Yuseong-gu, Daejeon 34141 (Korea, Republic of)

    2015-12-15

    Highlights: • A new PDHR strategy is proposed to cope with SBO-combined accidents. • The concept of integrated passive safety system (IPSS) is used in this strategy. • This strategy performs the functions of passive safety injection and SG gravity injection. • LOCAs in SBO are classified by the pressures in reactor coolant system for passive functions. • The strategy can be integrated with EOP and SAMG as a complementary strategy for ensuring safety. - Abstract: An integrated passive safety system (IPSS), to be achieved by the use of a large water tank placed at high elevation outside the containment, was proposed to achieve various passive functions. These include decay heat removal, safety injection, containment cooling, in-vessel retention through external reactor vessel cooling, and containment filtered venting. The purpose of the passive decay heat removal (PDHR) strategy using the IPSS is to cope with SBO and SBO-combined accidents under the assumption that existing engineered safety features have failed. In this paper, a PDHR strategy was developed based on the design and accident management strategy of Korean representative PWR, the OPR1000. The functions of a steam generator gravity injection and a passive safety injection system in the IPSS with safety depressurization systems were included in the PDHR strategy. Because the inadvertent opening of pressurizer valves and seal water leakage from RCPs could cause a loss of coolant in an SBO, LOCAs during a SBO were simulated to verify the performance of the strategy. The failure of active safety injection in LOCAs could also be covered by this strategy. Although LOCAs have generally been categorized according to their equivalent break diameters, the RCS pressure is used to classify the LOCAs during SBOs. The criteria values for categorization were determined from the proposed systems, which could maintain a reactor in a safe state by removing the decay heat for the SBO coping time of 8 h. The

  13. EC-sponsored research activities on innovative passive safety systems

    International Nuclear Information System (INIS)

    Bermejo, J.M.; Goethem, G. van

    2000-01-01

    On April 26th 1994, the European Union (EU) adopted via a Council Decision a EURATOM Multiannual Programme for community activities in the field of Nuclear Fission Safety (NFS) Research for the period 1994 to 1998. An area of work having, as an objective, to 'explore innovative approaches' to improve the safety of future and existing reactors, was introduced in this programme. Most of the projects selected in this area, which have been grouped under a common cluster known as 'INNO', are currently being carried out on a 'cost-shared' basis, i.e. contribution of the European Commission is up to 50% of the total cost. At present, the 'INNO' cluster is composed of 10 projects in which 25 different organisations, representing research centres, universities, regulators, utilities and vendors from 7 EU member states and Switzerland, are involved. These projects are proving to be an efficient means to gain the necessary phenomenological knowledge and to solve the challenging problems, many times of generic nature, posed among others by the characteristically small driving forces of the systems studied and by the lack of really prototypical test facilities. (author)

  14. Inherent/passive safety for fusion

    International Nuclear Information System (INIS)

    Piet, S.J.

    1986-06-01

    The concept of inherent or passive 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 safety and high mass power density, improving both safety and economics. When inherent safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be truly different and superior

  15. Survey of the passive safety systems of the BWR 1000 concept from SIEMENS

    Energy Technology Data Exchange (ETDEWEB)

    Mattern, J; Brettschuh, W; Palavecino, C [SIEMENS, Energieerzeugung, Offenbach (Germany)

    1996-12-01

    Through the use of passive safety systems and components for accident control in addition to the active systems required for plant operation, a higher degree of safety against core-endangering conditions is achieved which is no longer ruled by complex system engineering dependent on power supply and activation by I and C systems. A low core power density and large water inventories stored inside the reactor pressure vessel as well as inside and outside the containment ensure good plant behaviour in the event of transients or accidents. These passive safety systems - which required neither electric power to function nor I and C systems for actuation, being activated solely on the basis of changes in process variables such as water level, pressure and temperature - provide a grace period of more than 5 days after the onset of accident conditions before manual intervention becomes necessary. 8 figs.

  16. 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)

  17. Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

    International Nuclear Information System (INIS)

    Holcomb, David Eugene

    2015-01-01

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

  18. 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.

  19. Passive safety systems and natural circulation in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2009-11-01

    Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s 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 improve economics of new nuclear power plant designs. 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'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems 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. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

  20. A Regulatory Perspective on the Performance and Reliability of Nuclear Passive Safety Systems

    International Nuclear Information System (INIS)

    Quan, Pham Trung; Lee, Sukho

    2016-01-01

    Passive safety systems have been proven to enhance the safety of NPPs. When an accident such as station blackout occurs, these systems can perform the following functions: the decay heat removal, passive safety injection, containment cooling, and the retention of radioactive materials. Following the IAEA definitions, using passive safety systems reduces reliance on active components to achieve proper actuation and not requiring operator intervention in accident conditions. That leads to the deviations in boundary conditions of the critical process or geometric parameters, which activate and operate the system to perform accident prevention and mitigation functions. The main difficulties in evaluation of functional failure of passive systems arise because of (a) lack of plant operational experience; (b) scarcity of adequate experimental data from integral test facilities or from separate effect tests in order to understand the performance characteristics of these passive systems, not only at normal operation but also during accidents and transients; (c) lack of accepted definitions of failure modes for these systems; and (d) difficulty in modeling certain physical behavior of these systems. Reliability assessment of the PSS is still one of the important issues. Several reliability methodologies such as REPAS, RMPS and ASPRA have been applied to the reliability assessments. However, some issues are remained unresolved due to lack of understanding of the treatment of dynamic failure characteristics of components of the PSS, the treatment of dynamic variation of independence process parameters such as ambient temperature and the functional failure criteria of the PSS. Dynamic reliability methodologies should be integrated in the PSS reliability analysis to have a true estimate of system failure probability. The methodology should estimate the physical variation of the parameters and the frequency of the accident sequences when the dynamic effects are considered

  1. 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

  2. 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

  3. Assessment of passive safety system of a Small Modular Reactor (SMR)

    International Nuclear Information System (INIS)

    Butt, Hassan Nawaz; Ilyas, Muhammad; Ahmad, Masroor; Aydogan, Fatih

    2016-01-01

    Highlights: • The MASLWR test facility has been modeled in RELAP5-SCDAP. The model is validated by comparing the simulation results with the experimental data. • Results obtained from various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core. • New scenarios are considered in which the effect of vent and sump recirculation valves failure has been investigated. • It is found from the results that continuous loss of inventory occurs due to lack of recirculation. • It is concluded that the high pressure vent valves in the MASLWR safety system require more redundancy. - Abstract: Innovative SMRs are designed with enhanced safety features based on lessons learnt from past experience of plant operation. Reliance on natural circulation and addition of passive safety systems made them inherently safe and simple in design. It is required to study reliability assessment of passive safety systems during postulated transients prior to their deployment on commercial scale. Test facilities and best estimate system codes are playing significant role in assessment of passive safety systems as well as in design, certification and evaluation of these innovative types of reactors. RELAP5 code is widely used for thermal-hydraulic analysis of nuclear reactors. In this work, the passive safety systems of Multi-Application Small Light Water (MASLWR) have been assessed. The complete loop of the MASLWR test facility has been modeled in RELAP5-SCDAP Mod 4.0. The RELAP5 model is validated by comparing the simulation results with the experimental data. Results obtained for various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core to avoid core heat up. Some of the components of passive safety system of MASLWR still rely on active power. Therefore, it was necessary to investigate their performance under failure

  4. 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)

  5. Integral test facilities for validation of the performance of passive safety systems and natural circulation

    International Nuclear Information System (INIS)

    Choi, J. H.

    2010-10-01

    Passive safety systems are becoming an important component in advanced reactor designs. This has led to an international interest in examining natural circulation phenomena as this may play an important role in the operation of these passive safety systems. Understanding reactor system behaviour is a challenging process due to the complex interactions between components and associated phenomena. Properly scaled integral test facilities can be used to explore these complex interactions. In addition, system analysis computer codes can be used as predictive tools in understanding the complex reactor system behaviour. However, before the application of system analysis computer codes for reactor design, it is capability in making predictions needs to be validated against the experimental data from a properly scaled integral test facility. The IAEA has organized a coordinated research project (CRP) on natural circulation phenomena, modeling and reliability of passive systems that utilize natural circulation. This paper is a part of research results from this CRP and describes representative international integral test facilities that can be used for data collection for reactor types in which natural circulation may play an important role. Example experiments were described along with the analyses of these example cases in order to examine the ability of system codes to model the phenomena that are occurring in the test facilities. (Author)

  6. Transient simulation of ALWR passive safety systems using RELAP5/MOD2

    International Nuclear Information System (INIS)

    Elias, E.; Nekhamkin, Y.; Arshavski, I.

    2004-01-01

    Numerical simulation is presented of some passive safety systems currently incorporated in the design of the next generation advanced light water reactors (ALWRs). The performance and effectiveness of ex-core natural convection cooling and the concept of gravity driven water injection at high pressure are investigated using the RELAP5/MOD2 thermal-hydraulic code. The study identifies areas that should be investigated more fully in future experimental programs related to hypothetical large and small LOCA in ALWRs. (author)

  7. 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.

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

    International Nuclear Information System (INIS)

    2014-09-01

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

  9. Status and topics of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Ohnuki, Akira; Arai, Kenji; Kikuta, Michitaka; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime

    1999-01-01

    For increasing of electric power demand and reducing of carbon dioxide exhaust in the 21st century, studies of the next-generation light water reactor (LWR) with passive safety systems are developing in the world: AP-600 (by Westing House Co.); SBWR (by General Electric Co.); SWR1000 (by Siemens Co.); NP21 (by Mitsubishi Heavy Industry Co., et al.); JPSR (by JAERI). The passive equipment using natural circulation and natural convection are installed in the passive safety system, instead of active safety equipment, such as pumps, etc. It remains still as a important issue, however, to verify the reliability on the functions of the passive equipment, since that the driving forces of the passive equipment are small at comparison with the active safety equipment. The various subjects of thermal-hydraulic analysis for the next-generation light water reactors, such as temperature stratification in the passive safety systems, vapor condensation in the mixture of non-condensable gases and the interactions of the passive safety system with the primary cooling system, are illustrated and discussed in the paper. (M. Suetake)

  10. European vehicle passive safety network

    NARCIS (Netherlands)

    Wismans, J.S.H.M.; Janssen, E.G.

    1999-01-01

    The general objective of the European Vehicle Passive Safety Network is to contribute to the reduction of the number of road traffic victims in Europe by passive safety measures. The aim of the road safety policy of the European Commission is to reduce the annual total of fatalities to 18000 in

  11. 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

  12. 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).

  13. The REPAS approach to the evaluation of passive safety systems reliability

    International Nuclear Information System (INIS)

    Bianchi, F.; Burgazzi, L.; D'Auria, F.; Ricotti, M.E.

    2002-01-01

    Scope of this research, carried out by ENEA in collaboration with University of Pisa and Polytechnic of Milano since 1999, is the identification of a methodology allowing the evaluation of the reliability of passive systems as a whole, in a more physical and phenomenal way. The paper describe the study, named REPAS (Reliability Evaluation of Passive Safety systems), carried out by the partners and finalised to the development and validation of such a procedure. The strategy of engagement moves from the consideration that a passive system should be theoretically more reliable than an active one. In fact it does not need any external input or energy to operate and it relies only upon natural physical laws (e.g. gravity, natural circulation, internally stored energy, etc.) and/or 'intelligent' use of the energy inherently available in the system (e.g. chemical reaction, decay heat, etc.). Nevertheless the passive system may fail its mission not only as a consequence of classical mechanical failure of components, but also for deviation from the expected behaviour, due to physical phenomena mainly related to thermal-hydraulics or due to different boundary and initial conditions. The main sources of physical failure are identified and a probability of occurrence is assigned. The reliability analysis is performed on a passive system which operates in two-phase, natural circulation. The selected system is a loop including a heat source and a heat sink where the condensation occurs. The system behaviour under different configurations has been simulated via best-estimate code (Relap5 mod3.2). The results are shown and can be treated in such a way to give qualitative and quantitative information on the system reliability. Main routes of development of the methodology are also depicted. The analysis of the results shows that the procedure is suitable to evaluate the performance of a passive system on a probabilistic / deterministic basis. Important information can also be

  14. Some Findings from Thermal-Hydraulic Validation Tests for SMART Passive Safety System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Shin, Yong-Cheol; Min, Kyoung-Ho; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To satisfy the domestic and international needs for nuclear safety improvement after the Fukushima accident, an effort to improve its safety has been studied, and a Passive Safety System (PSS) for SMART has been designed. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMARTITL. Recently, a test program to validate the performance of the SMART PSS was launched and its scaled-down test facility was additionally installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests for the SMART PSS will be summarized. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide base data during the application phase of SDA revision and construction licensing. A test program to validate the performance of SMARS PSS was launched with an additional scaleddown test facility of SMART PSS, which will be installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests of the SMART passive safety system during 2013-2014 were summarized. They include a couple of SMART PSS tests using active pumps and several 1-train SMART PSS tests. From the test results it was estimated that the SMART PSS has sufficient cooling capability to deal with the SBLOCA scenario of SMART. During the SBLOCA scenario, in the CMT the water layer inventory was well stratified thermally and the safety injection water was injected efficiently into the RPV from the initial period and cools down the RCS properly.

  15. Information System Passive Safety at DaimlerChrysler: from vision to reality

    Energy Technology Data Exchange (ETDEWEB)

    Ruoff, H. [DaimlerChrysler AG, Sindelfingen (Germany)

    2001-07-01

    Today, short development times and an extensive model range are key factors when it comes to remaining a leading force in the highly competitive automobile market. Information technology plays a crucial role as an 'enabler' in keeping this competitive edge. It is only possible to make full use of existing potential by using 'state-of-the-art' information technology. In particular, this involves making knowledge that is distributed in various systems and 'heads' available quickly and transparently anytime and anywhere. The improved re-use of acquired expertise, efficient information searches, and standardized and seamless applications lead to a high level of performance. This often involves the transformation of a company culture from 'information hider' to 'information provider'. The vision of converting a gradually created, complex range of systems into a seamless sequence of distributed working processes with numerous participants is becoming reality in the Passive Safety area in Automobile/Vehicle Development at DaimlerChrysler. The INFOS Passive Safety information system provides all the data and information, which is required for, or results from, the entire crash test process. (orig.)

  16. Safety assessment for the passive system of the nuclear power plants (NPPs) using safety margin estimation

    International Nuclear Information System (INIS)

    Woo, Tae-Ho; Lee, Un-Chul

    2010-01-01

    The probabilistic safety assessment (PSA) for gas-cooled nuclear power plants has been investigated where the operational data are deficient, because there is not any commercial gas-cooled nuclear power plant. Therefore, it is necessary to use the statistical data for the basic event constructions. Several estimations for the safety margin are introduced for the quantification of the failure frequency in the basic event, which is made by the concept of the impact and affordability. Trend of probability of failure (TPF) and fuzzy converter (FC) are introduced using the safety margin, which shows the simplified and easy configurations for the event characteristics. The mass flow rate in the natural circulation is studied for the modeling. The potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are also investigated. The values in the probability set are compared with those of the fuzzy set modeling. Non-linearity of the safety margin is expressed by the fuzziness of the membership function. This artificial intelligence analysis of the fuzzy set could enhance the reliability of the system comparing to the probabilistic analysis.

  17. Assessment of Integrated Pedestrian Protection Systems with Autonomous Emergency Braking (AEB) and Passive Safety Components.

    Science.gov (United States)

    Edwards, Mervyn; Nathanson, Andrew; Carroll, Jolyon; Wisch, Marcus; Zander, Oliver; Lubbe, Nils

    2015-01-01

    Autonomous emergency braking (AEB) systems fitted to cars for pedestrians have been predicted to offer substantial benefit. On this basis, consumer rating programs-for example, the European New Car Assessment Programme (Euro NCAP)-are developing rating schemes to encourage fitment of these systems. One of the questions that needs to be answered to do this fully is how the assessment of the speed reduction offered by the AEB is integrated with the current assessment of the passive safety for mitigation of pedestrian injury. Ideally, this should be done on a benefit-related basis. The objective of this research was to develop a benefit-based methodology for assessment of integrated pedestrian protection systems with AEB and passive safety components. The method should include weighting procedures to ensure that it represents injury patterns from accident data and replicates an independently estimated benefit of AEB. A methodology has been developed to calculate the expected societal cost of pedestrian injuries, assuming that all pedestrians in the target population (i.e., pedestrians impacted by the front of a passenger car) are impacted by the car being assessed, taking into account the impact speed reduction offered by the car's AEB (if fitted) and the passive safety protection offered by the car's frontal structure. For rating purposes, the cost for the assessed car is normalized by comparing it to the cost calculated for a reference car. The speed reductions measured in AEB tests are used to determine the speed at which each pedestrian in the target population will be impacted. Injury probabilities for each impact are then calculated using the results from Euro NCAP pedestrian impactor tests and injury risk curves. These injury probabilities are converted into cost using "harm"-type costs for the body regions tested. These costs are weighted and summed. Weighting factors were determined using accident data from Germany and Great Britain and an independently

  18. Engineering reliability in design phase: An application to AP-600 reactor passive safety system

    International Nuclear Information System (INIS)

    Majumdr, D.; Siahpush, A.S.; Hills, S.W.

    1992-01-01

    A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

  19. 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

  20. CANDU passive shutdown systems

    Energy Technology Data Exchange (ETDEWEB)

    Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

    1996-12-01

    CANDU incorporates two diverse, passive shutdown systems, independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

  1. Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

    International Nuclear Information System (INIS)

    Morimoto, Y.; Madarame, H.; Okamoto, K.

    2001-01-01

    Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

  2. 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.

  3. MODULAR AND FULL SIZE SIMPLIFIED BOILING WATER REACTOR DESIGN WITH FULLY PASSIVE SAFETY SYSTEMS

    International Nuclear Information System (INIS)

    Ishii, M.; Revankar, S. T.; Downar, T.; Xu, Y.; Yoon, H. J.; Tinkler, D.; Rohatgi, U. S.

    2003-01-01

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  4. Fusion reactor passive safety and ignitor risk-based regulation

    International Nuclear Information System (INIS)

    Zucchetti, M.

    1995-01-01

    Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine

  5. 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)

  6. 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)

  7. 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

  8. Modeling Transients and Designing a Passive Safety System for a Nuclear Thermal Rocket Using Relap5

    Science.gov (United States)

    Khatry, Jivan

    Long-term high payload missions necessitate the need for nuclear space propulsion. Several nuclear reactor types were investigated by the Nuclear Engine for Rocket Vehicle Application (NERVA) program of National Aeronautics and Space Administration (NASA). Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. A NERVA design known as the Pewee I was selected for this purpose. The following transients were run: (i) modeling of corrosion-induced blockages on the peripheral fuel element coolant channels and their impact on radiation heat transfer in the core, and (ii) modeling of loss-of-flow-accidents (LOFAs) and their impact on radiation heat transfer in the core. For part (i), the radiation heat transfer rate of blocked channels increases while their neighbors' decreases. For part (ii), the core radiation heat transfer rate increases while the flow rate through the rocket system is decreased. However, the radiation heat transfer decreased while there was a complete LOFA. In this situation, the peripheral fuel element coolant channels handle the majority of the radiation heat transfer. Recognizing the LOFA as the most severe design basis accident, a passive safety system was designed in order to respond to such a transient. This design utilizes the already existing tie rod tubes and connects them to a radiator in a closed loop. Hence, this is basically a secondary loop. The size of the core is unchanged. During normal steady-state operation, this secondary loop keeps the moderator cool. Results show that the safety system is able to remove the decay heat and prevent the fuel elements from melting, in response to a LOFA and subsequent SCRAM.

  9. An experimental study on passive safety systems for the SMART design with the SMART-ITL facility

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yi, Sung-Jae

    2016-01-01

    Passive Safety Systems (PSSs) are added to the SMART design to increase the safety margin during accidents especially under a prolonged station blackout. A set of validation tests were performed for the PSSs of the SMART design with an integral effect test loop of SMART-ITL. Both single and dual trains of the Passive Safety Injection System (PSIS) were simulated to validate the SMART design together with two stages of Automatic Depressurization System (ADS) and four trains of Passive Residual Heat Removal System (PRHRS), and their results were compared. In this paper, the effect of the train number of PSIS on a Small-Break Loss of Coolant Accident (SBLOCA) scenario is investigated for a break size of 0.4 inch. The single and dual train tests show a similar trend in general but the injected water migrates slightly differently in the RV and is discharged through the break nozzle. The parameters of the Reactor Vessel (RV) pressure, RV water level, accumulated break mass, and injection flowrates from the Core Makeup Tank (CMT) and Safety Injection Tank (SIT) were compared. The acquired data will be used to validate the safety analysis code and its related models to evaluate the performance of SMART PSS, and to provide the base data during the application phase of construction licensing of the SMART design. (author)

  10. Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

    International Nuclear Information System (INIS)

    Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

    2014-01-01

    The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

  11. Testing of the multi-application small light water reactor (MASLWR) passive safety systems

    International Nuclear Information System (INIS)

    Reyes, Jose N.; Groome, John; Woods, Brian G.; Young, Eric; Abel, Kent; Yao, You; Yoo, Yeon Jong

    2007-01-01

    Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully

  12. A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

    International Nuclear Information System (INIS)

    Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

    1995-01-01

    The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

  13. Prediction of Heat Transfer Performance on Horizontal U-Shaped Heat Exchanger in Passive Safety System Using MARS

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech, Yongin (Korea, Republic of); Cho, Hyoung-Kyu; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    The design and the safety analysis of the passive safety systems are performed mainly using the best-estimate thermal-hydraulic analysis codes such as RELAP5 and MARS. This study developed the heat transfer model package for the horizontal U-shaped HX submerged in a pool by improving the horizontal in-tube condensation model and developing the outside-tube natural convective nucleate boiling model. This paper presents the HX model package and the validation results against the passive safety system-related experimental data of PASCAL and ATLAS-PAFS. This study developed the heat transfer model package of the horizontal U-shaped HX submerged in a pool in order to obtain a reliable prediction of the HX heat removal performance of the passive safety system, especially PAFS, using MARS. From the validation results, the proposed model package provided the improved prediction of HX performance (condensation, natural convective nucleate boiling, and heat removal rate of the HX) compared to the default model in MARS.

  14. 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

  15. Evaluation on driving force of natural circulation in downcomer for passive residual heat removal system in JAERI passive safety reactor JPSR

    International Nuclear Information System (INIS)

    Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

    1997-01-01

    The driving-force of the natural circulation in the residual heat removal (RHR) system for the JPSR (JAERI Passive Safety Reactor) is given as a gravity force of the density difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The amount of density difference and time to achieve the enough density difference for the RHR system change directly dependent on the thermal fluid flow pattern in downcomer of annulus flow pass. The purposes of the present study are to investigate the possibilities of the followings by evaluating the three-dimensional thermal fluid flow in downcomer by numerical analysis using the STREAM code; 1) promotion of making the flow pattern uniform in downcomer by installing a baffle, 2) achievement of an enough driving-force of the natural circulation, 3) validity of one-point assumption, that is, complete mixing down-flow assumption for the three-dimensional thermal fluid flow in downcomer to evaluate the function of the passive RHR system. The following conclusions were obtained: (1) The effect of baffle on the thermal fluid flow and driving-force is little, (2) The driving-force required for natural circulation cooling can be obtained in wide range of inlet velocity even if the flow is multi-dimensional, (3) Both in initial transient stage and in steady-state, the one-point assumption can be applied to evaluate the driving-force of natural circulation in the passive RHR system. (author)

  16. Major Results from 1-Train Passive Safety System Tests for the SMART Design with the SMART-ITL Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Ruy, Hyobong; Kim, Woo-Shik; Byun, Sun-Joon; Shin, Yong-Cheol; Min, Kyoung-Ho; Yi, Sung-Jae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    To satisfy the domestic and international needs for nuclear safety improvement after the Fukushima accident, an effort to improve its safety has been studied, and a Passive Safety System (PSS) for SMART has been designed. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMARTITL. Recently, a test program to validate the performance of the SMART PSS was launched and its scaled-down test facility was additionally installed at the existing SMART-ITL facility. In this paper, the major results from the 1-train passive safety system validation tests with the SMARTITL facility will be summarized. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide base data during the application phase of the SDA revision and construction licensing. In this paper, the major results from the validation tests of the SMART passive safety system using a 1-train test facility were summarized. They include a dozen of SMART PSS tests using 1-train SMART PSS tests. From the test results, it was estimated that the SMART PSS has sufficient cooling capability to deal with the SBLOCA scenario of SMART. During the SBLOCA scenario, in the CMT, the water layer inventory was well stratified thermally and the safety injection water was injected efficiently into the RPV from the initial period, and cools down the RCS properly.

  17. Concept of safety related I and C and power supply systems in the passive safety concept of the HTR-module

    International Nuclear Information System (INIS)

    Juengst, U.

    1990-01-01

    The main motivation for the passive safety concepts is to gain a better quality of safety or at least to achieve higher public acceptance for nuclear power plants. This strategy has been introduced into the European Fast Reactor (EER), a common project of France, UK and Germany is applied stringently to the German high-temperature gas-cooled reactor ''HTR - Module''. The following fields are briefly described in the paper: Safety design features of the HTR - Module, overview of I and C concept, reactor protection system, emergency control room, power supply concept, system arrangement and protection against external hazards, accidents sequence of station black-out. (author). 3 figs

  18. 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.

  19. System approach in the investigation of coolant parametrical oscillations in passive safety injection systems (PSIS)

    International Nuclear Information System (INIS)

    Proskouriakov, K.N.

    2001-01-01

    The use of thermal-hydraulic computer codes is an important part of the work programme for activities in the field of nuclear power plants (NPP) Safety Research as it will enable to define better the test configuration and parameter range extensions and to extrapolate the results of the small scale experiments towards full scale reactor applications. The CATHARE2, RELAP5, the WCOBRA/TRAC, and APROS codes are the estimate thermal hydraulic codes for the evaluation of large and small break loss of coolant accidents (LOCA). The relatively good agreement experimental data with the calculations have been presented. There was shown also some big mistakes in predicting distribution of flow when two phase are present. Model of parametrical oscillation (P.O.) worked out gives explanation for flow oscillations and indicates that the phenomenon of P.O. appears under certain combination of thermal-hydraulic parameters and structure of heat-removal system. (orig.)

  20. Westinghouse Advances in Passive Plant Safety

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  2. Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan

    2003-03-01

    This thesis is devoted to the investigation of passive safety and inherent features of subcritical nuclear transmutation systems - accelerator-driven systems. The general objective of this research has been to improve the safety performance and avoid elevated coolant temperatures in worst-case scenarios like unprotected loss-of-flow accidents, loss-of-heat-sink accidents, and a combination of both these accident initiators. The specific topics covered are emergency decay heat removal by reactor vessel auxiliary cooling systems, beam shut-off by a melt-rupture disc, safety aspects from locating heat-exchangers in the riser of a pool-type reactor system, and reduction of pressure resistance in the primary circuit by employing bypass routes. The initial part of the research was focused on reactor vessel auxiliary cooling systems. It was shown that an 80 MW{sub th} Pb/Bi-cooled accelerator-driven system of 8 m height and 6 m diameter vessel can be well cooled in the case of loss-of-flow accidents in which the accelerator proton beam is not switched off. After a loss-of-heat-sink accident the proton beam has to be interrupted within 40 minutes in order to avoid fast creep of the vessel. If a melt-rupture disc is included in the wall of the beam pipe, which breaks at 150 K above the normal core outlet temperature, the grace period until the beam has to be shut off is increased to 6 hours. For the same vessel geometry, but an operating power of 250 MW{sub th} the structural materials can still avoid fast creep in case the proton beam is shut off immediately. If beam shut-off is delayed, additional cooling methods are needed to increase the heat removal. Investigations were made on the filling of the gap between the guard and the reactor vessel with liquid metal coolant and using water spray cooling on the guard vessel surface. The second part of the thesis presents examinations regarding an accelerator-driven system also cooled with Pb/Bi but with heat-exchangers located

  3. Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

    International Nuclear Information System (INIS)

    Carlsson, Johan

    2003-03-01

    This thesis is devoted to the investigation of passive safety and inherent features of subcritical nuclear transmutation systems - accelerator-driven systems. The general objective of this research has been to improve the safety performance and avoid elevated coolant temperatures in worst-case scenarios like unprotected loss-of-flow accidents, loss-of-heat-sink accidents, and a combination of both these accident initiators. The specific topics covered are emergency decay heat removal by reactor vessel auxiliary cooling systems, beam shut-off by a melt-rupture disc, safety aspects from locating heat-exchangers in the riser of a pool-type reactor system, and reduction of pressure resistance in the primary circuit by employing bypass routes. The initial part of the research was focused on reactor vessel auxiliary cooling systems. It was shown that an 80 MW th Pb/Bi-cooled accelerator-driven system of 8 m height and 6 m diameter vessel can be well cooled in the case of loss-of-flow accidents in which the accelerator proton beam is not switched off. After a loss-of-heat-sink accident the proton beam has to be interrupted within 40 minutes in order to avoid fast creep of the vessel. If a melt-rupture disc is included in the wall of the beam pipe, which breaks at 150 K above the normal core outlet temperature, the grace period until the beam has to be shut off is increased to 6 hours. For the same vessel geometry, but an operating power of 250 MW th the structural materials can still avoid fast creep in case the proton beam is shut off immediately. If beam shut-off is delayed, additional cooling methods are needed to increase the heat removal. Investigations were made on the filling of the gap between the guard and the reactor vessel with liquid metal coolant and using water spray cooling on the guard vessel surface. The second part of the thesis presents examinations regarding an accelerator-driven system also cooled with Pb/Bi but with heat-exchangers located in the

  4. Passive safety device and internal short tested method for energy storage cells and systems

    Science.gov (United States)

    Keyser, Matthew; Darcy, Eric; Long, Dirk; Pesaran, Ahmad

    2015-09-22

    A passive safety device for an energy storage cell for positioning between two electrically conductive layers of the energy storage cell. The safety device also comprising a separator and a non-conductive layer. A first electrically conductive material is provided on the non-conductive layer. A first opening is formed through the separator between the first electrically conductive material and one of the electrically conductive layers of the energy storage device. A second electrically conductive material is provided adjacent the first electrically conductive material on the non-conductive layer, wherein a space is formed on the non-conductive layer between the first and second electrically conductive materials. A second opening is formed through the non-conductive layer between the second electrically conductive material and another of the electrically conductive layers of the energy storage device. The first and second electrically conductive materials combine and exit at least partially through the first and second openings to connect the two electrically conductive layers of the energy storage device at a predetermined temperature.

  5. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  6. 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

  7. Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

  8. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  9. Cost-benefit of the bubble tower concept as a containment passive safety system

    International Nuclear Information System (INIS)

    Iotti, R.C.; Bardach, H.; Shin, J.J.; Parnes, M.J.

    1994-01-01

    Containment system integrity for both PWRs and BWRs can be assured by passive measures highlighted the use of an accessory Bubble Tower. The utilization of the Bubble Tower precludes the possibility of containment overpressurization. From the thermodynamic standpoint, the Bubble Tower is simply water column of about 120 ft. height attached to the containment and connected to the air space above the suppression pool of a BWR, or a PWR In-containment Refueling Water Storage Tank. From the radiological protection standpoint, the Bubble Tower is a water column sufficient to effect decontamination factors of at least 100 for nuclide species other than the noble gases, and with the addition of organic solubilizers sufficient to effect decontamination factors of at least 10 iodides and at least 100 for other nuclide species. When containment steam or noncondensable gas passes through the Bubble Tower, a significant fraction of the radionuclides is absorbed by the water column. When a cost-benefit dose evaluation is performed relative to the utilization of a Bubble Tower, even under conditions where the dollars per man-rem is taken as $1000, the results are favorable. They are substantially more favorable when the dollars per man-rem is taken as $5000 or $10,000 as are the current trends. (author)

  10. 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

  11. 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.

  12. A Simple Fully Passive Safety Option for SMART SBLOCA

    International Nuclear Information System (INIS)

    Lee, Won Jae

    2012-01-01

    SMART reactor, an integral pressurized water reactor (iPWR), is developed by KAERI and now under standard design licensing review. Integral reactor design of the SMART has small diameter penetrations below 2 inches at upper parts of reactor pressure vessel (RPV) and the core is located at very lower part. Amount of reactor coolant inventory is around 0.55tons/MWth during normal operations, which is seven times more than that of conventional PWRs. Such intrinsic safety features of the SMART can provide prolonged core cooling during a small-break loss-of-coolant accident (SBLOCA). As an engineered safety feature for SBLOCA, electrically two-train and mechanically four-train active safety injection (SI) systems are provided to refill the RPV, whose safety been proven through safety analysis and experiments. In addition, four-train passive residual heat removal systems (PRHRSs) are provided to remove core decay heat by natural circulation in the secondary side of steam generators during transient and accident conditions. After Fukushima disaster, a passive safety of nuclear power plants has become more emphasized than conventional active safety, even though there are still debates whether it can really insure the realistic safety. Passive safety is defined such that the core safety is ensured for 72 hours after accidents without any active safety systems and operator actions. In light of this, a simple fully passive safety option for SBLOCA is proposed: low-pressure safety injection tanks (SITs) and heat pipes submerged in the PRHRS emergency coolant tanks (ECTs). Post-LOCA long-term cooling after 72 hours is provided by sump recirculation using shutdown cooling system. Realistic analysis method using MARS3.1 is used to derive fully passive safety option, and then to screen design and operating parameters and to demonstrate the safety performance of SITs. SI line break is selected as a reference SBLOCA scenario

  13. 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

  14. Expanding pedestrian injury risk to the body region level: how to model passive safety systems in pedestrian injury risk functions.

    Science.gov (United States)

    Niebuhr, Tobias; Junge, Mirko; Achmus, Stefanie

    2015-01-01

    decomposable into the 3 body regions and so are the risk functions as body region-specific risk functions. The risk functions for each body region are stated explicitly for different injury severity levels and compared to the real-world accident data. The body region-specific risk functions can then be used to model the effect of improved passive safety systems. These modified body region-specific injury risk functions are aggregated to a new pedestrian injury risk function. Passive safety systems can therefore be modeled in injury risk functions for the first time. A short example on how the results can be used for assessing the effectiveness of new driver assistance systems concludes the article.

  15. Development of IAEA description of passive safety and subsequent thoughts

    Energy Technology Data Exchange (ETDEWEB)

    Lang, P M [USDOE, Washington, DC (United States)

    1996-12-01

    The description of passive components and systems published by the IAEA in its TECDOC-626 was developed in the course of a Technical Committee Meeting held in Sweden and two subsequent Consultants Meetings held in Vienna. This description is reviewed and discussed in terms of the philosophies behind it, alternatives considered, problems encountered, and conclusions drawn. Also discussed is an Appendix to the TECDOC, which illustrates the spectrum of possibilities from passive to active by describing four typical categories of passivity. Subsequent thoughts on passive safety include a discussion of its advantages and disadvantages, concluding with a summary of current views and problems with it. (author). 8 refs.

  16. Specialists' meeting on passive and active safety features of LMFRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-07-01

    The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics.

  17. Specialists' meeting on passive and active safety features of LMFRs

    International Nuclear Information System (INIS)

    1991-01-01

    The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  19. 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.

  20. 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

  1. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    International Nuclear Information System (INIS)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

    1997-01-01

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

  2. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  3. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

    1998-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  4. 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

  5. [Injuries caused by traffic accidents: passive safety and restraint systems in automobiles].

    Science.gov (United States)

    Zuppichini, F; Orlandi, E; Genna, M; Rodella, L; Ricci, G; Arienzo, A; Dorrucci, V; Inaspettato, G

    1986-10-01

    In this article are considered the multiple instruments today employed in cars, in order to prevent or ameliorate the lesions caused to the occupants in case of road accident. The acquisitions in the differentiated structure of the car, in the windshield, in the components of the passenger cell are described, and the peculiar importance of the restraint systems is evidenced.

  6. 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

  7. 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.

  8. Integration of the functional reliability of two passive safety systems to mitigate a SBLOCA+BO in a CAREM-like reactor PSA

    Energy Technology Data Exchange (ETDEWEB)

    Mezio, Federico, E-mail: federico.mezio@cab.cnea.gov.ar [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Grinberg, Mariela [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina); Lorenzo, Gabriel [CNEA, Sede Central, Av. Del Libertador 8250, CABA (Argentina); Giménez, Marcelo [CNEA, Centro Atómico Bariloche, S.C. de Bariloche, Río Negro (Argentina)

    2014-04-01

    Highlights: • An estimation of the Functional Unreliability was performed using RMPS methodology. • The methodology uses an improved response surface in order to estimate the FU. • The FU may become relevant to be analyzed in the Passive Safety Systems. • There were proposed two ways to incorporate the FU into an APS. - Abstract: This paper describes a case study of a methodological approach for assessing the functional reliability of passive safety systems (PSS) and its treatment within a probabilistic safety assessment (PSA). The functional unreliability (FU) can be understood as the failure probability of PSS to fulfill its mission due to the impairment of the related passive safety function. The safety function accomplishment is characterized and quantified by a performance indicator (PI), which is a measure of how far the system is from verifying its mission. PI uncertainties are estimated from uncertainty propagation of selected parameters. A methodology based on the reliability methodology for passive system (RMPS) one is used to estimate the FU associated to the isolation condensers (ICs) in combination with the accumulators (medium pressure injection system) of a CAREM-like integral advanced reactor. A small break loss of coolant accident with black-out is selected as an evaluation case. This implies success of reactor shut-down (inherent) and failure of residual heat removal by active systems. The safety function to accomplish is to refill the reactor pressure vessel (RPV) in order to avoid core damage. For this case, to allow the discharge of accumulators into RPV, the pressure must be reduced by the IC. The methodology for passive safety function assessment considers uncertainties in code parameters, besides uncertainties in engineering parameters (design, construction, operation and maintenance), in order to perform Monte Carlo simulations based on best estimate (B-E) plant model. Then, response surfaces based on PI are used for improving the

  9. Basic investigation of particle swarm optimization performance in a reduced scale PWR passive safety system design

    International Nuclear Information System (INIS)

    Cunha, Joao J. da; Lapa, Celso Marcelo F.; Alvim, Antonio Carlos M.; Lima, Carlos A. Souza; Pereira, Claudio Marcio do N.A.

    2010-01-01

    This work presents a methodology to investigate the viability of using particle swarm optimization technique to obtain the best combination of physical and operational parameters that lead to the best adjusted dimensionless groups, calculated by similarity laws, that are able to simulate the most relevant physical phenomena in single-phase flow under natural circulation and to offer an appropriate alternative reduced scale design for reactor primary loops with this flow characteristics. A PWR reactor core, under natural circulation, based on LOFT test facility, was used as the case study. The particle swarm optimization technique was applied to a problem with these thermo-hydraulics conditions and results demonstrated the viability and adequacy of the method to design similar systems with these characteristics.

  10. Basic investigation of particle swarm optimization performance in a reduced scale PWR passive safety system design

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear, Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7o andar. Centro, Rio de Janeiro 20091-906 (Brazil); Lapa, Celso Marcelo F., E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, P.O. Box 68509, Cidade Universitaria, Ilha do Fundao s/n, Rio de Janeiro 21945-970 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil); Lima, Carlos A. Souza [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel, s/n, Vila Nova, Nova Friburgo 28630-050 (Brazil); Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear, Divisao de Reatores/PPGIEN, P.O. Box 68550, Rua Helio de Almeida 75 Cidade Universitaria, Ilha do Fundao, Rio de Janeiro 21941-972 (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (Brazil)

    2010-03-15

    This work presents a methodology to investigate the viability of using particle swarm optimization technique to obtain the best combination of physical and operational parameters that lead to the best adjusted dimensionless groups, calculated by similarity laws, that are able to simulate the most relevant physical phenomena in single-phase flow under natural circulation and to offer an appropriate alternative reduced scale design for reactor primary loops with this flow characteristics. A PWR reactor core, under natural circulation, based on LOFT test facility, was used as the case study. The particle swarm optimization technique was applied to a problem with these thermo-hydraulics conditions and results demonstrated the viability and adequacy of the method to design similar systems with these characteristics.

  11. 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

  12. Use of RMPS to assess the reliability of Passive Safety Systems in CAREM-like reactor, past and present experiences. Second progress report

    International Nuclear Information System (INIS)

    Giménez, M; Mezio, F.; Zanocco, P.; Lorenzo, G.

    2011-01-01

    Conclusions: • RMPS is being used successfully to assess the fulfillment of design criteria from a probabilistic point of view, in case of LOHS and LOCA, considering uncertainties in the reactor, in the passive safety systems and in the models as well. • Allows to quantify the probability of Event Tree headers related to some systems whose demand depends on the accidental sequence evolution (i.e. probability to demand a safety valve in case of a LOHS with success of the PRHRS, but working under deteriorated conditions). • Functional reliability quantification not already used in CAREM PSA, (Fault Trees or in Event Trees?)

  13. Establishment of design concept of large capacity passive reactor KP1000 and performance evaluation of safety system for LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O.; Hwang, Young Dong; Kim, Young In; Chang, Moon Hee

    1997-03-01

    This study was performed to establish the design concepts and to evaluate the performance of safety features of large capacity passive reactor (1000 MWe grade). The design concepts of the large capacity passive reactor `KP1000` were established to generate 1000 MW electric power based on the AP600 of Westinghouse by increasing the number of reactor coolant loop and by increasing the size of reactor internals/core. To implement the analysis of the LBLOCA for KP1000, various kinds of computer codes being considered, it was concluded that RELAP5 was the most appropriate one in availability and operations in present situation. By the analysis of the computer code `RELAP5/Mod3.2.1.2`, following conclusions were derived as described below. First, by spectrum analysis of the discharge factor of the berak part, the most conservative discharge factor C{sub D}=1.2 and the PCT value of KP1000 was 1254F, which is slightly higher than the value of AP600 but is much less than the existing active reactor `Kori 3 and 4` where blowdown PCT value is 1693.4 deg F and reflooding PCT is 1918.4 deg F. Second, after the 200 seconds from the initiation of LBLOCA, IRWST water was supplied in a stable state and the maximum temperature of clad were maintained in a saturated condition. Therefore, it was concluded that the passive safety features of KP1000 keep reactor core from being damaged for large break LOCA. (author). 11 refs., 28 tabs., 37 figs.

  14. Status of IAEA CRPI31018 “Development of Methodologies for the Assessment of Passive Safety System Performance in Advanced Reactors”

    International Nuclear Information System (INIS)

    Subki, Hadid M.

    2011-01-01

    Purpose of research coordination meeting: • To review progress and milestones on all research activities; • To discuss the preliminary experimental data obtained from the Natural Circulation Loop Facility L2 in Italy constructed for the assessment of different methodologies for the evaluation of the reliability of passive safety system; • To discuss lessons-to be-learned from the Fukushima Daiichi Accident in Japan and its implications to near future R&D needs on thermal-hydraulics and reactor safety; • To develop an outline of integrated annual technical report and future collaboration plan

  15. Third (3rd) Research Coordination Meeting of the CRP on Development of Methodologies for the Assessment of Passive Safety System Performance in Advanced Reactors. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    Purpose of the meeting: • To review progress and milestones on all research activities; • To discuss the preliminary experimental data obtained from the Natural Circulation Loop Facility L2 in Italy constructed for the assessment of different methodologies for the evaluation of the reliability of passive safety system; • To discuss lessons-to be-learned from the Fukushima Daiichi Accident in Japan and its implications to near future R&D needs on thermal-hydraulics and reactor safety; • To develop an outline of integrated annual technical report and future collaboration plan

  16. 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)

  17. Passive magnetic bearing system

    Science.gov (United States)

    Post, Richard F.

    2014-09-02

    An axial stabilizer for the rotor of a magnetic bearing provides external control of stiffness through switching in external inductances. External control also allows the stabilizer to become a part of a passive/active magnetic bearing system that requires no external source of power and no position sensor. Stabilizers for displacements transverse to the axis of rotation are provided that require only a single cylindrical Halbach array in its operation, and thus are especially suited for use in high rotation speed applications, such as flywheel energy storage systems. The elimination of the need of an inner cylindrical array solves the difficult mechanical problem of supplying support against centrifugal forces for the magnets of that array. Compensation is provided for the temperature variation of the strength of the magnetic fields of the permanent magnets in the levitating magnet arrays.

  18. 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

  19. Passive safety injection experiments and analyses (PAHKO)

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1998-01-01

    PAHKO project involved experiments on the PACTEL facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines (Pressure Balancing Line, PBL, and Injection Line, IL). The examined PSIS worked efficiently in SBLOCAs although the flow through the PSIS stopped temporarily if the break was very small and the hot water filled the CMT. The experiments demonstrated the importance of the flow distributor in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable to simulate the overall behaviour of the transients. The detailed analyses of the results showed some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of the PSIS phenomena. (orig.)

  20. 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

  1. Safety analysis of increase in heat removal from reactor coolant system with inadvertent operation of passive residual heat removal at no load conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shao, Ge; Cao, Xuewu [School of Mechanical and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-06-15

    The advanced passive pressurized water reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) system was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary coolant system, the actuation of the PRHR will enhance the cooldown of the primary coolant system. There is a risk of power excursion during the cooldown of the primary coolant system. Therefore, it is necessary to analyze the thermal hydraulic behavior of the reactor coolant system (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or reactor coolant system in the selected conditions.

  2. 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

  3. Passive and engineered safety features of the prototype fast reactor (PFR), Dounreay

    International Nuclear Information System (INIS)

    Gregory, C.V.

    1991-01-01

    Prototype fast reactor (PFR) combines passive and engineered safety features. Natural convection, a strong negative power coefficient, the decay heat removal system, and a fuel design able to operate beyond failure are all inherent and passive safety features of the PFR. The reliable shutdown system and the protection provided against SGU leaks are example of engineered protection. Experience at PFR demonstrates the worth and potential of a range of passive and engineered safeguards

  4. Italy: Analysis of Solutions for Passively Actuated Safety Shutdown Devices

    International Nuclear Information System (INIS)

    Burgazzi, L.

    2015-01-01

    This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on Fast Reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in Sodium Fast Reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to Fast Reactor design to meet the safety requirements

  5. 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

  6. 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.

  7. Divergent effects of transformational and passive leadership on employee safety.

    Science.gov (United States)

    Kelloway, E Kevin; Mullen, Jane; Francis, Lori

    2006-01-01

    The authors concurrently examined the impact of safety-specific transformational leadership and safety-specific passive leadership on safety outcomes. First, the authors demonstrated via confirmatory factor analysis that safety-specific transformational leadership and safety-specific passive leadership are empirically distinct constructs. Second, using hierarchical regression, the authors illustrated, contrary to a stated corollary of transformational leadership theory (B. M. Bass, 1997), that passive leadership contributes incrementally to the prediction of organizationally relevant outcomes, in this case safety-related variables, beyond transformational leadership alone. Third, further analyses via structural equation modeling showed that both transformational and passive leadership have opposite effects on safety climate and safety consciousness, and these variables, in turn, predict safety events and injuries. Implications for research and application are discussed. Copyright 2006 APA.

  8. NucleDyne's passive containment system

    International Nuclear Information System (INIS)

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

    1987-01-01

    A simple definition of the passive containment system is that it is a total safeguards system for light water reactors designed to prevent and contain any accidental release of radioactivity. Its passive features utilize the natural laws of physics and thermodynamics. The system encompasses three basic containments constructed as one integrated structure on the reactor building foundation. The primary containment encloses the reactor pressure vessel and coolant system and passive engineered safety systems and components. Auxiliary containment enclosures house auxiliary systems and components. Secondary containment (the reactor building), housing the primary and auxiliary containment structures, provides a second containment barrier as added defense-in-depth against leakage of radioactivity for all accidents assumed by the industry. The generic features of the passive containment system are applicable to both the boiling water reactors and the pressurized water reactors as standardized features for all power ranges. These features provide for a zero source term, the industry's ultimate safety goal. This paper relates to a four-loop pressurized water reactor

  9. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  10. 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.

  11. 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)

  12. Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

    Energy Technology Data Exchange (ETDEWEB)

    Tuunanen, J. [VTT Energy, Espoo (Finland). Nuclear Energy; Vihavainen, J. [Lappeenranta Univ. of Technology (Finland); D' Auria, F. [Univ. of Pisa (Italy); Kimber, G. [AEA Technology (United Kingdom)

    1999-07-01

    The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

  13. Assessment of passive safety injection systems of ALWRs. Final report of the European Commission 4th framework programme. Project FI4I-CT95-004 (APSI)

    International Nuclear Information System (INIS)

    Tuunanen, J.; D'Auria, F.; Kimber, G.

    1999-01-01

    The European Commission 4th Framework Programme project 'Assessment of Passive Safety Injection Systems of Advanced Light Water Reactors (FI4I-CT95-0004)' involved experiments on the PACTEL test facility and computer simulations of selected experiments. The experiments focused on the performance of Passive Safety Injection Systems (PSIS) of Advanced Light Water Reactors (ALWRs) in Small Break Loss-Of-Coolant Accident (SBLOCA) conditions. The PSIS consisted of a Core Make-up Tank (CMT) and two pipelines. A pressure balancing line (PBL) connected the CMT to one cold leg. The injection line (IL) connected it to the downcomer. The project involved 15 experiments in three series. The experiments provided valuable information about condensation and heat transfer processes in the CMT, thermal stratification of water in the CMT, and natural circulation flow through the PSIS lines. The experiments showed the examined PSIS works efficiently in SBLOCAs although the flow through the PSIS may stop in very small SBLOCAs, when the hot water fills the CMT. The experiments also demonstrated the importance of flow distributor (sparger) in the CMT to limit rapid condensation. The project included validation of three thermal-hydraulic computer codes (APROS, CATHARE and RELAP5). The analyses showed the codes are capable of simulating the overall behaviour of the transients. The codes predicted accurately the core heatup, which occurred when the primary coolant inventory was reduced so much that the core top became free of water. The detailed analyses of the calculation results showed that some models in the codes still need improvements. Especially, further development of models for thermal stratification, condensation and natural circulation flow with small driving forces would be necessary for accurate simulation of phenomena in the PSIS. (orig.)

  14. Comparative Investigation on 0.4 inch SBLOCA Scenario with Single and Dual Train Passive Safety Injection Systems using SMART-ITL

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yun, Eun-Koo; Choi, Nam-Hyun; Min, Kyoung-Ho; Shin, Yong-Cheol; Bang, Yoon-Gon; Kim, Myoung-Jun; Seo, Chan-Jong; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Standard Design Approval (SDA) for SMART was certificated in 2012 at the Korea Atomic Energy Research Institute (KAERI). In December 2015, Saudi Arabia and Korea started conducting a three-year project of Pre-Project Engineering (PPE) to prepare a Preliminary Safety Analysis Report (PSAR) and to review the feasibility of constructing SMART reactors in Saudi Arabia. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMART-ITL. In this paper, a comparative investigation was performed on 0.4 inch SBLOCA scenario with single and dual train passive safety injection systems using SMART-ITL. In this paper, the effect of the train number of PSIS on a SBLOCA scenario is investigated for a break size of 0.4 inch. The single and dual train tests show a similar trend in general but the injected water migrates slightly differently in the RV and is discharged through the break nozzle. The parameters of the RV pressure, RV water level, accumulated break mass, and injection flowrates from the CMT and SIT were compared. Compared with the single train test, the increased injection rates from the two trains of the PSIS during the dual train test raised the RV water level, ensuring the safety of the reactor core.

  15. Analysis for Passive Safety Injection of IPSS in Various LOCAs

    International Nuclear Information System (INIS)

    Kim, Sangho; Chang, Soonheung

    2013-01-01

    The Fukushima accident shows US the possibility of accidents that are beyond a designed imagination. Lots of lessons can be shortly summarized into three issues. First of all, the original cause was the occurrence of a Station Black-Out (SBO). Even if engineers considered the possibility of a loss of offsite power enough to be managed, the failure of EDGs seemed to be unnoticed. The second is poor operation and accident management. They could not understand the overall system and did not check the availability of alternating systems. The third is the large release of radioactive materials outside the containment. Even if SBO occurred and the accident was not managed well, all the means must have prevented the large release out of containment. After that, lots of problems were pointed and numerous actions were carried out in each country. The representative proposals are AAC, additional physical barrier, bunker concept and large big tank. Integrated passive safety system (IPSS) was proposed as one of the solutions for enhancing the safety. IPSS can cope with a SBO and accidents with a SBO. IPSS has five functions which are passive decay heat removal, passive safety injection, passive containment cooling, passive in-vessel retention and filtered venting system. The results showed a high performance of removing decay heat through steam generator cooling by forming natural circulation in the primary circuit. The design concept of passive safety injection system (PSIS) consists of the injection line from integrated passive safety tank (IPST) to reactor vessel. The previous works were only focused on a double ended guillotine break LOCA in SBO. The purpose of this paper is to analyze the performance of PSIS in IPSS for various LOCAs by using MARS (Multi-dimensional Analysis of Reactor Safety) code. The simulated accidents were LOCAs which were accompanied with a SBO. The conditions of the LOCAs were varied only for the size of break. It shall show the capability of PSIS

  16. The Evaluation of the Safety Benefits of Combined Passive and On-Board Active Safety Applications

    Science.gov (United States)

    Page, Yves; Cuny, Sophie; Zangmeister, Tobias; Kreiss, Jens-Peter; Hermitte, Thierry

    2009-01-01

    One of the objectives of the European TRACE project (TRaffic Accident Causation in Europe, 2006–2008) was to estimate the proportion of injury accidents that could be avoided and/or the proportion of injury accidents where the severity could be mitigated for on-the-market safety applications, if 100 % of the car fleet would be equipped with them. We have selected for evaluation the Electronic Stability Control (ESC) and the Emergency Brake Assist (EBA) applications. As for passive safety systems, recent cars are designed to offer overall safety protection. Car structure, load limiters, front airbags, side airbags, knee airbags, pretensioners, padding and non aggressive structures in the door panel, the dashboard, the windshield, the seats, and the head rest also contribute to applying more protection. The whole safety package is very difficult to evaluate separately, one element independently segmented from the others. We decided to consider evaluating the effectivenessof the whole passive safety package, This package,, for the sake of simplicity, was the number of stars awarded at the Euro NCAP testing. The challenges were to compare the effectiveness of some safety configuration SC I, with the effectiveness of a different safety configuration SC II. A safety configuration is understood as a package of safety functions. Ten comparisons have been carried out such as the evaluation of the safety benefit of a fifth star given that the car has four stars and an EBA. The main outcome of this analysis is that any addition of a passive or active safety function selected in this analysis is producing increased safety benefits. For example, if all cars were five stars fitted with EBA and ESC, instead of four stars without ESC and EBA, injury accidents would be reduced by 47.2% for severe injuries and 69.5% for fatal injuries. PMID:20184838

  17. 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)

  18. Passive safety design characteristics of the KALIMER-600 burner reactor

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

    2009-01-01

    The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

  19. Some important issues in evaluating the availability of passive systems

    International Nuclear Information System (INIS)

    Kafka, P.; Kelemen, I.; Krzykacz, B.

    1993-01-01

    In some countries new reactor concepts based on a broader use of passive safety features are under development. The term 'passive' as used here refers to systems which rely heavily on natural heat transfer process such as natural circulation to perform their function rather than on decidedly 'active' components like pumps. The paper deals with important issues in evaluating the availability of passive systems, e.g. the assessment of the active components, the assessment of passive components and structures and the probabilistic assessment of the physical function of the natural processes. Based on an outlined assessment process for the entire system and on an exercised simulation process for the assessment of passive components e.g. pipes insights and important issues are presented. A follow-up study will refine and expand the concept to a full scope assessment procedure of passive systems. (author)

  20. A study of passive safety conditions for fast reactor core

    International Nuclear Information System (INIS)

    Shimizu, Akinao

    1991-01-01

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

  1. 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.

  2. Transformational and passive leadership as cross-level moderators of the relationships between safety knowledge, safety motivation, and safety participation.

    Science.gov (United States)

    Jiang, Lixin; Probst, Tahira M

    2016-06-01

    While safety knowledge and safety motivation are well-established predictors of safety participation, less is known about the impact of leadership styles on these relationships. The purpose of the current study was to examine whether the positive relationships between safety knowledge and motivation and safety participation are contingent on transformational and passive forms of safety leadership. Using multilevel modeling with a sample of 171 employees nested in 40 workgroups, we found that transformational safety leadership strengthened the safety knowledge-participation relationship, whereas passive leadership weakened the safety motivation-participation relationship. Under low transformational leadership, safety motivation was not related to safety participation; under high passive leadership, safety knowledge was not related to safety participation. These results are discussed in light of organizational efforts to increase safety-related citizenship behaviors. Copyright © 2016 Elsevier Ltd and National Safety Council. All rights reserved.

  3. 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

  4. 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.

  5. Germany: Assessment of the efficiency of a passive safety system for prevention of severe accidents for SFR

    International Nuclear Information System (INIS)

    Bubelis, E.

    2015-01-01

    The aim of the study was the evaluation of severe transient behavior in Sodium-cooled Fast Reactor (SFR) and of the impact of newly conceived inherent mitigation measures (the use of ASD – additional shutdown device). The SFR design taken for the analysis was the SFR(v2b-ST) reactor design, and the system code to be used was selected to be the SIM-SFR code. The transients chosen for evaluation of the efficiency of mitigation measures were the unprotected loss-of-flow (ULOF) and the unprotected loss-of-heat-sink (ULOHS)

  6. Solar Passive Modification Increase Radiation Safety Standards Inside Accelerator Building

    International Nuclear Information System (INIS)

    Eid, A. F.; Keshk, A. B.

    2010-01-01

    Irradiation processing by accelerated electrons is considering one of the most important and useful industrial irradiation treatments. It is depending on two principle attachment elements which are architecture of irradiation building and the accelerator characteristic that was arranged inside irradiation building. Negative environmental measurements were recorded inside the main building and were exceeded the international standards (humidity, air speed, high thermal effects and ozone concentration). The study showed that it is essential to improve the natural environmental standards inside the main irradiation building in order to improve the work environment and to reduce ozone concentration from 220 ppb to international standard. The main goals and advantages were achieved by using environmental architecture (desert architecture) indoor the irradiation building. The work depends on passive solar system which is economic, same architectural elements, comfort / health, and radiation safety, and without mechanical means. The experimental work was accomplished under these modifications. The registered results of various environmental concentrations have proved their normal standards.

  7. 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)

  8. 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

  9. Hybrid Active-Passive Radiation Shielding System

    Data.gov (United States)

    National Aeronautics and Space Administration — A radiation shielding system is proposed that integrates active magnetic fields with passive shielding materials. The objective is to increase the shielding...

  10. Passive and inherent safety technologies for light-water nuclear reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

  11. Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

    International Nuclear Information System (INIS)

    Araya, F.

    1995-01-01

    After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

  12. 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

  13. 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)

  14. Safety Systems

    Science.gov (United States)

    Halligan, Tom

    2009-01-01

    Colleges across the country are rising to the task by implementing safety programs, response strategies, and technologies intended to create a secure environment for teachers and students. Whether it is preparing and responding to a natural disaster, health emergency, or act of violence, more schools are making campus safety a top priority. At…

  15. 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)

  16. Feasibility study on emergency passive habitability systems of SPWR

    International Nuclear Information System (INIS)

    Obata, H.; Tabata, H.; Urakami, M.; Naito, T.

    2000-01-01

    The major characteristic of the Simplified Pressurized Water Reactor (SPWR) is that safety systems for the emergency core cooling and the core decay heat removal functions are achieved by passive equipment. The AP600 developed in the U.S adopts passive emergency habitability system for the main control room (MCR) and the electrical equipment rooms (EER) by using the concrete of the structures as a heat sink. For the SPWR, alternative natural circulation cooling systems have been investigated: for MCR cooling, a cold water reservoir is used as heat sink; for EER cooling, outside air is instead employed. The distribution of the air-velocity and temperature in those rooms were calculated by using a three-dimensional thermal fluid analysis code. The authors verified the conceptual feasibility of these systems as the emergency passive habitability systems in the SPWR. (author)

  17. 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

  18. Analysis of solutions for passively activated safety shutdown devices for SFR

    International Nuclear Information System (INIS)

    Burgazzi, Luciano

    2013-01-01

    Highlights: • Innovative systems for emergency shut down of fast reactors are proposed. • The concepts of inherent and passive safety are put forward. • The relative analysis in terms of safety and reliability is presented. • A comparative assessment among the concepts is performed. • Path forward is tracked. -- Abstract: In order to enhance the inherent safety of fast reactors, innovative reactivity control systems have been proposed for intrinsic ultimate shut-down instead of conventional scram rods, to cope with the potential consequences of severe unprotected transient accidents, such as an energetic core disruptive accident, as in case of sodium fast reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to fast reactor design to meet the safety requirements. This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on fast reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in sodium fast reactors

  19. 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

  20. 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.

  1. Thermal-hydraulic unreliability of passive systems

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Saltos, N.T.

    1995-01-01

    Advanced light water reactor designs like AP600 and the simplified boiling water reactor (SBWR) use passive safety systems for accident prevention and mitigation. Because these systems rely on natural forces for their operation, their unavailability due to hardware failures and human error is significantly smaller than that of active systems. However, the coolant flows predicted to be delivered by these systems can be subject to significant uncertainties, which in turn can lead to a significant uncertainty in the predicted thermal-hydraulic performance of the plant under accident conditions. Because of these uncertainties, there is a probability that an accident sequence for which a best estimate thermal-hydraulic analysis predicts no core damage (success sequence) may actually lead to core damage. For brevity, this probability will be called thermal-hydraulic unreliability. The assessment of this unreliability for all the success sequences requires very expensive computations. Moreover, the computational cost increases drastically as the required thermal-hydraulic reliability increases. The required computational effort can be greatly reduced if a bounding approach can be used that either eliminates the need to compute thermal-hydraulic unreliabilities, or it leads to the analysis of a few bounding sequences for which the required thermal-hydraulic reliability is relatively small. The objective of this paper is to present such an approach and determine the order of magnitude of the thermal-hydraulic unreliabilities that may have to be computed

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

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

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

  3. Active and passive fatigue in simulated driving: discriminating styles of workload regulation and their safety impacts.

    Science.gov (United States)

    Saxby, Dyani J; Matthews, Gerald; Warm, Joel S; Hitchcock, Edward M; Neubauer, Catherine

    2013-12-01

    Despite the known dangers of driver fatigue, it is a difficult construct to study empirically. Different forms of task-induced fatigue may differ in their effects on driver performance and safety. Desmond and Hancock (2001) defined active and passive fatigue states that reflect different styles of workload regulation. In 2 driving simulator studies we investigated the multidimensional subjective states and safety outcomes associated with active and passive fatigue. Wind gusts were used to induce active fatigue, and full vehicle automation to induce passive fatigue. Drive duration was independently manipulated to track the development of fatigue states over time. Participants were undergraduate students. Study 1 (N = 108) focused on subjective response and associated cognitive stress processes, while Study 2 (N = 168) tested fatigue effects on vehicle control and alertness. In both studies the 2 fatigue manipulations produced different patterns of subjective response reflecting different styles of workload regulation, appraisal, and coping. Active fatigue was associated with distress, overload, and heightened coping efforts, whereas passive fatigue corresponded to large-magnitude declines in task engagement, cognitive underload, and reduced challenge appraisal. Study 2 showed that only passive fatigue reduced alertness, operationalized as speed of braking and steering responses to an emergency event. Passive fatigue also increased crash probability, but did not affect a measure of vehicle control. Findings support theories that see fatigue as an outcome of strategies for managing workload. The distinction between active and passive fatigue is important for assessment of fatigue and for evaluating automated driving systems which may induce dangerous levels of passive fatigue. PsycINFO Database Record (c) 2013 APA, all rights reserved.

  4. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  5. Passive safety testing at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Lucoff, D.M.

    1989-01-01

    During 1986, the Fast Flux Test Facility (FFTF) conducted several tests designed to improve the understanding of the passive safety characteristics of an oxide-fueled liquid-metal reactor (LMR). Static and dynamic tests were performed over a broad range of power, flow, and temperature conditions that extended beyond those for normal operation. Key results of these tests are presented. Stable operation at low power with natural circulation cooling was demonstrated. A passive safety enhancement feature, the gas expansion module (GEM) was developed specifically to offset the large amount of cooldown reactivity that needs to be controlled in an oxide-fueled LMR undergoing an unprotected loss-of-flow accident. Nine GEMs were built and successfully tested in FFTF. With the reactor at 50% power (200 MW (thermal)), the main coolant pumps were turned off and the normal control rod scram response was inhibited. The GEMs and inherent core reactivity feedback mechanisms took the core subcritical with a modest peak coolant temperature transient that reached 85 degrees C above the pretransient value and always maintained a >400 degrees C margin to the sodium boiling point (910 degrees C)

  6. Chaotic behavior in a system simulating the pressure balanced injection system. Analysis of passive safety reactor behavior. JAERI's nuclear research promotion program, H12-012 (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Madarame, Haruki; Okamoto, Koji; Tanaka, Gentaro; Morimoto, Yuichiro [Tokyo Univ., School of Engineering, Tokyo (Japan); Sato, Akira [Yamagata Univ., Faculty of Engineering, Yonezawa, Yamagata (Japan); Kondou, Masaya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The pressure Balanced Injection System (PBIS) was proposed in a passive safety reactor. Pressurizing Line (PL) connects the Reactor Vessel (RV) and the gas area in the Contain Vessel (CV), and Injected Line (IL) connects two vessels at relatively lower position. In an accident, the two lines are passively opened. The vapor generated by the residual heat pressed downward the water level in the RV. When the level is lower than the inlet of the PL, vapor is ejected into the CV through the PL attaining the pressure balance between the vessels. Then boron water in the CV is injected into the RV through the IL by the static head. This process is repeated by the succeeding vapor generation. In an experiment, the oscillating system was replaced by water column in a U-shaped duct. The vapor generation was simulated by cover gas supply to one end of the duct, while the other end was open to the atmosphere. When the water level reached a certain level, electromagnetic valves opened and the cover gas was ejected. The gas pressure decreased rapidly, resulting in a surface rise. When the water level reached another level, the valves closed. The cover gas pressure increased again, thus, gas ejection occurred intermittently. The interval of the gas ejection was not constant but fluctuated widely. Mere stochastic noise could hardly explain the large amplitude. Then was expressed the system using a set of linear equations. Various types of piecewise linear model were developed to examine the cause of the fluctuation. There appeared tangential bifurcation, period-doubling bifurcation, period-adding bifurcation and so on. The calculated interval exhibited chaotic features. Thus the cause of the fluctuation can be attributed to chaotic features of the system having switching. Since the piecewise linear model was highly simplified the behavior, a quantitative comparison between the calculation and the experiment was difficult. Therefore, numerical simulation code considering nonlinear

  7. 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

  8. 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

  9. A Review: Passive System Reliability Analysis – Accomplishments and Unresolved Issues

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, Arun Kumar, E-mail: arunths@barc.gov.in [Reactor Engineering Division, Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai (India); Chandrakar, Amit [Homi Bhabha National Institute, Mumbai (India); Vinod, Gopika [Reactor Safety Division, Reactor Design and Development Group, Bhabha Atomic Research Centre, Mumbai (India)

    2014-10-10

    Reliability assessment of passive safety systems is one of the important issues, since safety of advanced nuclear reactors rely on several passive features. In this context, a few methodologies such as reliability evaluation of passive safety system (REPAS), reliability methods for passive safety functions (RMPS), and analysis of passive systems reliability (APSRA) have been developed in the past. These methodologies have been used to assess reliability of various passive safety systems. While these methodologies have certain features in common, but they differ in considering certain issues; for example, treatment of model uncertainties, deviation of geometric, and process parameters from their nominal values. This paper presents the state of the art on passive system reliability assessment methodologies, the accomplishments, and remaining issues. In this review, three critical issues pertaining to passive systems performance and reliability have been identified. The first issue is applicability of best estimate codes and model uncertainty. The best estimate codes based phenomenological simulations of natural convection passive systems could have significant amount of uncertainties, these uncertainties must be incorporated in appropriate manner in the performance and reliability analysis of such systems. The second issue is the treatment of dynamic failure characteristics of components of passive systems. REPAS, RMPS, and APSRA methodologies do not consider dynamic failures of components or process, which may have strong influence on the failure of passive systems. The influence of dynamic failure characteristics of components on system failure probability is presented with the help of a dynamic reliability methodology based on Monte Carlo simulation. The analysis of a benchmark problem of Hold-up tank shows the error in failure probability estimation by not considering the dynamism of components. It is thus suggested that dynamic reliability methodologies must be

  10. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  11. AC-600 passive ECRHR system and its research program

    International Nuclear Information System (INIS)

    Chen Bingde; Xiao Zejun; Zhou Renmin; Liu Yiyang

    1997-01-01

    The secondary-side passive emergency core residual heat removal system (ECRHR System) is an important part of AC-600 PWR passive safety system, with which the core decay heat can be removed through nature circulation in primary and secondary system. Since 1991, the program for AC-600 passive ECRHR system has been conducted to investigate its distinct thermal-hydraulic phenomena, heat removal capability, affecting factors, and to develop computer codes. The test facility, designed according to the power/volume simulating law, is a full pressure and temperature operating loop with volume scaling factor of 1/390. It is composed of main loop system, emergence feedwater system, depression system, heat tracing, I and C system and power supply system. A total of sixteen tests is planned in first stage and fifteen of them have been done. The preliminary result analysis showed that the system has efficient heat removal capability in most conditions and some special thermal hydraulic phenomena, for example, flow fluctuation, which has negative impact on system's nature circulation, were identified

  12. Inherently safe passive gas monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Cordaro, Joseph V.; Bellamy, John Stephen; Shuler, James M.; Shull, Davis J.; Leduc, Daniel R.

    2016-09-06

    Generally, the present disclosure is directed to gas monitoring systems that use inductive power transfer to safely power an electrically passive device included within a nuclear material storage container. In particular, the electrically passive device can include an inductive power receiver for receiving inductive power transfer through a wall of the nuclear material storage container. The power received by the inductive power receiver can be used to power one or more sensors included in the device. Thus, the device is not required to include active power generation components such as, for example, a battery, that increase the risk of a spark igniting flammable gases within the container.

  13. 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

  14. SAFETY OF PASSIVE HOUSES SUBJECTED TO EARTHQUAKE, FINAL REPORT

    Directory of Open Access Journals (Sweden)

    Vojko Kilar

    2013-12-01

    Full Text Available he topic researched within the applied project. "Safety of passive houses subjected to earthquake" stemmed from two otherwise quite unrelated fields, i.e. seismic resistance and energy efficiency that in European countries do not frequently appear together. Just in Slovenia these two fields join each other, so identifying the problem and establishment of research right in Slovenia represents uniqueness and specificity. The majority of Slovenia is situated in area of moderate seismic risk. In order to ensure adequate mechanical resistance and stability of structures constructed in such area, the consideration of seismic effects is required by law. In Slovenia the number of passive houses and energy efficient buildings increases rapidly. However, for the time being the structural solutions that have been developed and broadly applied mainly in the areas with low seismicity (where the structural control to vertical static loads is sufficient are used. In earthquake-prone areas also adequate resistance to dynamic seismic effects have to be assured.

  15. 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)

  16. Development of passive radar systems at TNO

    NARCIS (Netherlands)

    Gelsema, S.J.

    2007-01-01

    Since 2002, the Netherlands Organisation for Applied Scientific Research – TNO, has been involved in the development of passive radar systems for research purposes. The development has been sponsored partly by the Royal Netherlands Air Force – whose main interest is threat evaluation – and partly by

  17. Plant experience with check valves in passive systems

    Energy Technology Data Exchange (ETDEWEB)

    Pahladsingh, R R [GKN Joint Nuclear Power Plant, Dodewaard (Netherlands)

    1996-12-01

    In the design of the advanced nuclear reactors there is a tendency to introduce more passive safety systems. The 25 year old design of the GKN nuclear reactor is different from the present BWR reactors because of some special features, such as the Natural Circulation - and the Passive Isolation Condenser system. When reviewing the design, one can conclude that the plant has 25 years of experience with check valves in passive systems and as passive components in systems. The result of this experience has been modeled in a plant-specific ``living PSA`` for the plant. A data-analysis has been performed on components which are related to the safety systems in the plant. As part of this study also the check valves have been taken in consideration. At GKN, the check valves have shown to be reliable components in the systems and no catastrophic failures have been experienced during the 25 years of operation. Especially the Isolation Condenser with its operation experience can contribute substantially to the insight of check valves in stand-by position at reactor pressure and operating by gravity under different pressure conditions. With the introduction of several passive systems in the SBWR-600 design, such as the Isolation Condensers, Gravity Driven Cooling, and Suppression Pool Cooling System, the issue of reliability of check valves in these systems is actual. Some critical aspects for study in connection with check valves are: What is the reliability of a check valve in a system at reactor pressure, to open on demand; what is the reliability of a check valve in a system at low pressure (gravity), to open on demand; what is the reliability of a check valve to open/close when the stand-by check wave is at zero differential pressure. The plant experience with check valves in a few essential safety systems is described and a brief introduction will be made about the application of check valves in the design of the new generation reactors is given. (author). 6 figs, 1 tab.

  18. GOTHIC Simulation of Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Ha, Huiun; Kim, Hangon

    2013-01-01

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR

  19. GOTHIC Simulation of Passive Containment Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Kim, Hangon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR.

  20. Backup passive reactivity shutdown systems

    International Nuclear Information System (INIS)

    Ashurko, Yu.M.; Kuznetsov, L.A.

    1996-01-01

    The paper reviews self-actuated shutdown systems (SASSs) for liquid metal-cooled fast reactors (LMFRs). Principles of operation are described, advantages and drawbacks analyzed, and prospects for application in advanced fast reactors examined. Ways to improve reactor self-protection via reactivity feedback amplification and related problems are discussed. (author). 9 refs, 12 figs

  1. Backup passive reactivity shutdown systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

    The paper reviews self-actuated shutdown systems (SASSs) for liquid metal-cooled fast reactors (LMFRs). Principles of operation are described, advantages and drawbacks analyzed, and prospects for application in advanced fast reactors examined. Ways to improve reactor self-protection via reactivity feedback amplification and related problems are discussed. (author). 9 refs, 12 figs.

  2. System safety education focused on flight safety

    Science.gov (United States)

    Holt, E.

    1971-01-01

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

  3. COMMIX analysis of AP-600 Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

    1992-01-01

    COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

  4. Active and passive vehicle safety at Volkswagen accident research

    Energy Technology Data Exchange (ETDEWEB)

    Jungmichel, M.; Stanzel, M.; Zobel, R. [Volkswagen AG, Wolfsburg (Germany)

    2001-07-01

    Accident Analysis is an efficient means of improving vehicle passive safety and is used frequently and intensively. However, reliable data on accident causation is much more difficult to obtain. In most cases, one or more of the persons involved in an accident will face litigation and therefore are reluctant to provide the information that is essential to researchers. In addition, antilock brakes in almost every current vehicle have caused certain characteristic evidence, i.e. skid marks, to appear much less frequently than before. However, this evidence provides valuable information for assessing the reaction of the driver and his attempt to avoid the accident. In order to implement strategies of accident avoidance, accident causation must first be fully understood. Therefore, one of the assignments of the Volkswagen Accident Research Unit is to interpret global statistics, as well as to study single cases in order to come up with strategies for collision avoidance or mitigation. Currently, our primary concern is focused on active vehicle safety by researching vehicle behavior in the pre-crash phase. (orig.)

  5. Integration of Active and Passive Safety Technologies--A Method to Study and Estimate Field Capability.

    Science.gov (United States)

    Hu, Jingwen; Flannagan, Carol A; Bao, Shan; McCoy, Robert W; Siasoco, Kevin M; Barbat, Saeed

    2015-11-01

    The objective of this study is to develop a method that uses a combination of field data analysis, naturalistic driving data analysis, and computational simulations to explore the potential injury reduction capabilities of integrating passive and active safety systems in frontal impact conditions. For the purposes of this study, the active safety system is actually a driver assist (DA) feature that has the potential to reduce delta-V prior to a crash, in frontal or other crash scenarios. A field data analysis was first conducted to estimate the delta-V distribution change based on an assumption of 20% crash avoidance resulting from a pre-crash braking DA feature. Analysis of changes in driver head location during 470 hard braking events in a naturalistic driving study found that drivers' head positions were mostly in the center position before the braking onset, while the percentage of time drivers leaning forward or backward increased significantly after the braking onset. Parametric studies with a total of 4800 MADYMO simulations showed that both delta-V and occupant pre-crash posture had pronounced effects on occupant injury risks and on the optimal restraint designs. By combining the results for the delta-V and head position distribution changes, a weighted average of injury risk reduction of 17% and 48% was predicted by the 50th percentile Anthropomorphic Test Device (ATD) model and human body model, respectively, with the assumption that the restraint system can adapt to the specific delta-V and pre-crash posture. This study demonstrated the potential for further reducing occupant injury risk in frontal crashes by the integration of a passive safety system with a DA feature. Future analyses considering more vehicle models, various crash conditions, and variations of occupant characteristics, such as age, gender, weight, and height, are necessary to further investigate the potential capability of integrating passive and DA or active safety systems.

  6. Sum rules and constraints on passive systems

    International Nuclear Information System (INIS)

    Bernland, A; Gustafsson, M; Luger, A

    2011-01-01

    A passive system is one that cannot produce energy, a property that naturally poses constraints on the system. A system in convolution form is fully described by its transfer function, and the class of Herglotz functions, holomorphic functions mapping the open upper half-plane to the closed upper half-plane, is closely related to the transfer functions of passive systems. Following a well-known representation theorem, Herglotz functions can be represented by means of positive measures on the real line. This fact is exploited in this paper in order to rigorously prove a set of integral identities for Herglotz functions that relate weighted integrals of the function to its asymptotic expansions at the origin and infinity. The integral identities are the core of a general approach introduced here to derive sum rules and physical limitations on various passive physical systems. Although similar approaches have previously been applied to a wide range of specific applications, this paper is the first to deliver a general procedure together with the necessary proofs. This procedure is described thoroughly and exemplified with examples from electromagnetic theory.

  7. Reactor system safety assurance

    International Nuclear Information System (INIS)

    Mattson, R.J.

    1984-01-01

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

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

    International Nuclear Information System (INIS)

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2014-01-01

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

  10. Investigating Car Body Construction Influence on the Passive Safety in a Rear Impact

    Directory of Open Access Journals (Sweden)

    D. Yu. Solopov

    2015-01-01

    Full Text Available This article solves the task to assess how much a car body construction influences on the parameters of passive safety in a rear impact. The task concerns the impact and, as a result, is highly nonlinear with large values of deformations, stresses, and accelerations. A finite element method based on software systems LS-DYNA, ANSYS, FEMAP, and others solves this task.One of the most important stages of the work was to develop the finite element models (FEM of the car as a whole, as well as the car seat with a dummy mounted in the car. Body of the Chiseler Grand Caravan car, which parameters are close to average ones, was used as an object of research.The results of calculations and experiments allowed us to find that in assessing the passive safety of a car, taking into consideration the body design with a seat mounted in it, values of velocities, accelerations, and NIC criterion turned out to be lower than when calculating the seat with a dummy separately. The relative error (relative to the results of calculations in the "dangerous" impact of FEM seat of the highest level in accordance with EURO NCAP was 32% for full acceleration and was 33% for NIC criterion.It was found that in the calculations based on the FEM car, as a whole, the results are more accurate than when using the load operation conditions simulating energy absorption by the car body (20%.This leads to the conclusion that the calculations based on the FEM car with the seat mounted in it gives the possibility to design a seat (with passive or active headrest to ensure the best level of passive safety of this car.

  11. Passive sensor systems for nuclear material monitoring

    International Nuclear Information System (INIS)

    Simpson, M.L.; Boatner, L.A.; Holcomb, D.E.; McElhaney, S.A.; Mihalczo, J.T.; Muhs, J.D.; Roberts, M.R.; Hill, N.W.

    1993-01-01

    Passive fiber optic sensor systems capable of confirming the presence of special nuclear materials in storage or process facilities are being developed at Oak Ridge National Laboratory (ORNL). These sensors provide completely passive, remote measurement capability. No power supplies, amplifiers, or other active components that could degrade system reliability are required at the sensor location. ORNL, through its research programs in scintillator materials, has developed a variety of materials for use in alpha-, beta-, gamma-, and neutron-sensitive scintillator detectors. In addition to sensors for measuring radiation flux, new sensor materials have been developed which are capable of measuring weight, temperature, and source location. An example of a passive sensor for temperature measurement is the combination of a thermophosphor (e.g., rare-earth activated Y 2 O 3 ) with 6 LiF (95% 6 Li). This combination results in a new class of scintillators for thermal neutrons that absorb energy from the radiation particles and remit the energy as a light pulse, the decay rate of which, over a specified temperature range, is temperature dependent. Other passive sensors being developed include pressure-sensitive triboluminescent materials, weight-sensitive silicone rubber fibers, scintillating fibers, and other materials for gamma and neutron detection. The light from the scintillator materials of each sensor would be sent through optical fibers to a monitoring station, where the attribute quantity could be measured and compared with previously recorded emission levels. Confirmatory measurement applications of these technologies are being evaluated to reduce the effort, costs, and employee exposures associated with inventorying stockpiles of highly enriched uranium at the Oak Ridge Y-12 Plant

  12. 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.)

  13. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  14. Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

    2004-06-01

    The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

  15. 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

  16. Evaluation of the nucledyne passive containment system

    International Nuclear Information System (INIS)

    1981-04-01

    This reports contains: (1) an evaluation by Gilbert/Commonwealth (G/C) of the NucleDyne passive Containment System (PCS) as that conceptual design is applied to a Westinghouse, two loop, Pressurized Water Reactor; (2) an evaluation by Westinghouse of two questions about the impact of the PCS on the Nuclear Steam Supply System (NSSS), which were posed by G/C and best answered by an NSSS vendor; and (3) replies to both the Gilbert/Commonwealth report and the Westinghoue report by NucleDyne Engineering Corporation

  17. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  18. 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.

  19. 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.

  20. Safety system function trends

    International Nuclear Information System (INIS)

    Johnson, C.

    1989-01-01

    This paper describes research to develop risk-based indicators of plant safety performance. One measure of the safety-performance of operating nuclear power plants is the unavailability of important safety systems. Brookhaven National Laboratory and Science Applications International Corporation are evaluating ways to aggregate train-level or component-level data to provide such an indicator. This type of indicator would respond to changes in plant safety margins faster than the currently used indicator of safety system unavailability (i.e., safety system failures reported in licensee event reports). Trends in the proposed indicator would be one indication of trends in plant safety performance and maintenance effectiveness. This paper summarizes the basis for such an indicator, identifies technical issues to be resolved, and illustrates the potential usefullness of such indicators by means of computer simulations and case studies

  1. Experimental and design experience with passive safety features of liquid metal reactors

    International Nuclear Information System (INIS)

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

  2. A Reliability Assessment Method for the VHTR Safety Systems

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  3. Design and transient analyses of passive emergency feedwater system of CPR1000. Part 1. Air cooling condition

    International Nuclear Information System (INIS)

    Zhang Yapei; Qiu Suizheng; Su Guanghui; Tian Wenxi; Cao Jianhua; Lu Donghua; Fu Xiangang

    2011-01-01

    The steam generator secondary passive emergency feedwater system is a new design for traditional generation Ⅱ + reactor CPR1000. The passive emergency feedwater system is designed to supply water to the SG shell side and improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The passive emergency feedwater system consists of steam generator (SG), heat exchanger (HX), air cooling tower, emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, the model of the primary loop system and the passive emergency feedwater system was developed to investigate residual heat removal capability of the passive emergency feedwater system and the transient characteristics of the primary loop system affected by the passive emergency feedwater system using RELAP5/MOD3.4. The transient characteristics of the primary loop system and the passive emergency feedwater system were calculated in the event of feed line break accident. Sensitivity studies of the passive emergency feedwater system were also conducted to investigate the response of the primary loop and the passive emergency feedwater system on the main parameters of the passive emergency feedwater system. The passive emergency feedwater system could supply water to the SG shell side from the EMT successfully. The calculation results showed that the passive emergency feedwater system could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and passive emergency feedwater system loop, respectively. (author)

  4. THE USE OF PASSIVE SOLAR HEATING SYSTEMS AS PART OF THE PASSIVE HOUSE

    Directory of Open Access Journals (Sweden)

    Bryzgalin Vladislav Viktorovich

    2018-05-01

    Full Text Available Subject: systems of passive solar heating, which can, without the use of engineering equipment, capture and accumulate the solar heat used for heating buildings. Research objectives: study of the possibility to reach the passive house standard (buildings with near zero energy consumption for heating in climatic conditions of Russia using the systems of passive solar heating in combination with other solutions for reduction of energy costs of building developed in the past. Materials and methods: search and analysis of literature, containing descriptions of various passive solar heating systems, examples of their use in different climatic conditions and the resulting effect obtained from their use; analysis of thermophysical processes occurring in these systems. Results: we revealed the potential of using the solar heating systems in the climatic conditions of parts of the territories of the Russian Federation, identified the possibility of cheaper construction by the passive house standard with the use of these systems. Conclusions: more detailed analysis of thermophysical and other processes that take place in passive solar heating systems is required for creation of their computational models, which will allow us to more accurately predict their effectiveness and seek the most cost-effective design solutions, and include them in the list of means for achieving the passive house standard.

  5. An experimental study of the flow characteristics of fluidic device in a passive safety injection tank

    International Nuclear Information System (INIS)

    Cho, Seok; Song, Chul Hwa; Won, Suon Yeon; Min, Kyong Ho; Chung, Moon Ki

    1998-01-01

    It is considered to adopt passive safety injection tank (SIT) as a enhanced safety feature in KNGR. Passive SIT employs a vortex chamber as a fluidic device, which control injection flow rate passively by the variation of flow resistance produced by vortex intensity within the vortex chamber. To investigate the flow characteristics of the vortex chamber many tests have been carried out by using small-scale test facility. In this report the effects of geometric parameters of vortex chamber on discharge flow characteristics and the velocity measurement result of flow field, measured by PIV, are presented and discussed. (author). 25 refs., 11 tabs., 31 figs

  6. Plant control impact on IFR power plant passive safety response

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1993-01-01

    A method is described for optimizing the closed-loop plant control strategy with respect to safety margins sustained in the unprotected upset response of a liquid metal reactor. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  7. 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

  8. 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

  9. Technical Meeting on Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2015-01-01

    A major focus of the design of modern fast reactor systems is on inherent and passive safety. Specific systems to improve reactor safety performance during accidental transients have been developed in nearly all fast reactor programs, and a large number of proposed systems have reached various stages of maturity. This Technical Meeting on Passive Shutdown Systems for Fast Reactors, which was recommended by the Technical Working Group on Fast Reactors (TWG-FR), addressed Member States’ expressed need for information exchange on projects and programs in the field, as well as for the identification of priorities based on the analysis of technology gaps to be covered through R&D activities. This meeting was limited to shutdown systems only, and did not include other passive features such as natural circulation decay heat removal systems etc.; however the meeting catered to passive shutdown safety devices applicable to all types of fast neutron systems. It was agreed to initiate a new study and produce a Nuclear Energy Series (NES) Technical Report to collect information about the existing operational systems as well as innovative concepts under development. This will be a useful source for member states interested in gaining technical expertise to develop passive shutdown systems as well as to highlight the importance and development in this area

  10. Sampled Data Systems Passivity and Discrete Port-Hamiltonian Systems

    NARCIS (Netherlands)

    Stramigioli, Stefano; Secchi, Cristian; Schaft, Arjan J. van der; Fantuzzi, Cesare

    2005-01-01

    In this paper, we present a novel way to approach the interconnection of a continuous and a discrete time physical system. This is done in a way which preserves passivity of the coupled system independently of the sampling time T. This strategy can be used both in the field of telemanipulation, for

  11. Safety system status monitoring

    International Nuclear Information System (INIS)

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide

  12. Safety system status monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.

  13. Performance of the PBX-M passive plate stabilization system

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, R.; Bernabei, S.

    1994-02-01

    The PBX-M passive plate stabilization system provides significant stabilization of long-wavelength external kink modes, the slowing of vertical instability growth rates, and the amelioration of disruption characteristics. The passive plate stabilization system has allowed the use of LHCD and IBW to induce current density and pressure profile modifications, and m = 1 divertor biasing for modifying edge plasma transport. Improvements in the passive plate system insulators and support structures have provided reliable operation. Impurity influxes with the close-fitting passive plates are low. Solid target boronization is applied routinely to reduce conditioning time and maintain clean conditions

  14. 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

  15. 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.

  16. 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.

  17. Reactor safety systems

    International Nuclear Information System (INIS)

    Kafka, P.

    1975-01-01

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

  18. Safety design requirements for safety systems and components of JSFR

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

    2011-01-01

    Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

  19. Screening key parameters related to passive system performance based on Analytic Hierarchy Process

    International Nuclear Information System (INIS)

    Ma, Guohang; Yu, Yu; Huang, Xiong; Peng, Yuan; Ma, Nan; Shan, Zuhua; Niu, Fenglei; Wang, Shengfei

    2015-01-01

    Highlights: • An improved AHP method is presented for screening key parameters used in passive system reliability analysis. • We take the special bottom parameters as criterion for calculation and the abrupt change of the results are verified. • Combination weights are also affected by uncertainty of input parameters. - Abstract: Passive safety system is widely used in the new generation nuclear power plant (NPP) designs such as AP1000 to improve the reactor safety benefitting from its simple construction and less request for human intervene. However, the functional failure induced by uncertainty in the system thermal–hydraulic (T–H) performance becomes one of the main contributors to system operational failure since the system operates based on natural circulation, which should be considered in the system reliability evaluation. In order to improve the calculation efficiency the key parameters which significantly affect the system T–H characteristics can be screened and then be analyzed in detail. The Analytical Hierarchy Process (AHP) is one of the efficient methods to analyze the influence of the parameters on a passive system based on the experts’ experience. The passive containment cooling system (PCCS) in AP1000 is one of the typical passive safety systems, nevertheless too many parameters need to be analyzed and the T–H model itself is more complicated, so the traditional AHP method should be mended to use for screening key parameters efficiently. In this paper, we adapt the improved method in hierarchy construction and experts’ opinions integration, some parameters at the bottom justly in the traditional hierarchy are studied as criterion layer in improved AHP, the rationality of the method and the effect of abrupt change with the data are verified. The passive containment cooling system (PCCS) in AP1000 is evaluated as an example, and four key parameters are selected from 49 inputs

  20. Final report-passive safety optimization in liquid sodium-cooled reactors

    International Nuclear Information System (INIS)

    Cahalana, J. E.; Hahn, D.

    2007-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  1. Incremental passivity and output regulation for switched nonlinear systems

    Science.gov (United States)

    Pang, Hongbo; Zhao, Jun

    2017-10-01

    This paper studies incremental passivity and global output regulation for switched nonlinear systems, whose subsystems are not required to be incrementally passive. A concept of incremental passivity for switched systems is put forward. First, a switched system is rendered incrementally passive by the design of a state-dependent switching law. Second, the feedback incremental passification is achieved by the design of a state-dependent switching law and a set of state feedback controllers. Finally, we show that once the incremental passivity for switched nonlinear systems is assured, the output regulation problem is solved by the design of global nonlinear regulator controllers comprising two components: the steady-state control and the linear output feedback stabilising controllers, even though the problem for none of subsystems is solvable. Two examples are presented to illustrate the effectiveness of the proposed approach.

  2. Passive versus active hazard detection and avoidance systems

    Science.gov (United States)

    Neveu, D.; Mercier, G.; Hamel, J.-F.; Simard Bilodeau, V.; Woicke, S.; Alger, M.; Beaudette, D.

    2015-06-01

    Upcoming planetary exploration missions will require advanced guidance, navigation and control technologies to reach landing sites with high precision and safety. Various technologies are currently in development to meet that goal. Some technologies rely on passive sensors and benefit from the low mass and power of such solutions while others rely on active sensors and benefit from an improved robustness and accuracy. This paper presents two different hazard detection and avoidance (HDA) system design approaches. The first architecture relies only on a camera as the passive HDA sensor while the second relies, in addition, on a Lidar as the active HDA sensor. Both options use in common an innovative hazard map fusion algorithm aiming at identifying the safest landing locations. This paper presents the simulation tools and reports the closed-loop software simulation results obtained using each design option. The paper also reports the Monte Carlo simulation campaign that was used to assess the robustness of each design option. The performance of each design option is compared against each other in terms of performance criteria such as percentage of success, mean distance to nearest hazard, etc. The applicability of each design option to planetary exploration missions is also discussed.

  3. Software system safety

    Science.gov (United States)

    Uber, James G.

    1988-01-01

    Software itself is not hazardous, but since software and hardware share common interfaces there is an opportunity for software to create hazards. Further, these software systems are complex, and proven methods for the design, analysis, and measurement of software safety are not yet available. Some past software failures, future NASA software trends, software engineering methods, and tools and techniques for various software safety analyses are reviewed. Recommendations to NASA are made based on this review.

  4. Passive safety optimization in liquid-sodium cooled reactors

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hahn, D.; Chang, W.-P.; Kwon, Y.-M.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2004-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4)

  5. 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

  6. Passive neutron interrogation in systems with a poorly characterized detection efficiency

    International Nuclear Information System (INIS)

    Dubi, Chen; Oster, Elad; Ocherashvilli, Aharon; Pedersen, Bent; Hutszy, Janus

    2014-01-01

    Passive neutron interrogation for fissile mass estimation, relying on neutrons coming from spontaneous fission events, is considered a standard NDT procedure in the nuclear safeguard and safety community. Since most structure materials are (relatively) transparent to neutron radiation, passive neutron interrogation is considered highly effective in the analysis of dirty, poorly characterized samples. On the other hand, since a typical passive interrogation assembly is based on 3He detectors, neutrons from additional neutron sources (mainly (α,n) reactions and induced fissions in the tested sample) cannot be separated from the main spontaneous fission source through energetic spectral analysis. There for, applying the passive interrogation methods the implementation of Neutron Multiplicity Counting (NMC) methods for separation between the main fission source and the additional sources. Applying NMC methods requires a well characterized system, in the sense that both system die away time and detection efficiency must be well known (and in particular, independent of the tested sample)

  7. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  8. Reactor safety assessment system

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Safety Information System Guide

    International Nuclear Information System (INIS)

    Bullock, M.G.

    1977-03-01

    This Guide provides guidelines for the design and evaluation of a working safety information system. For the relatively few safety professionals who have already adopted computer-based programs, this Guide may aid them in the evaluation of their present system. To those who intend to develop an information system, it will, hopefully, inspire new thinking and encourage steps towards systems safety management. For the line manager who is working where the action is, this Guide may provide insight on the importance of accident facts as a tool for moving ideas up the communication ladder where they will be heard and acted upon; where what he has to say will influence beneficial changes among those who plan and control his operations. In the design of a safety information system, it is suggested that the safety manager make friends with a computer expert or someone on the management team who has some feeling for, and understanding of, the art of information storage and retrieval as a new and better means for communication

  10. Certification of passive houses : Lessons from real indoor climate systems

    NARCIS (Netherlands)

    Mlecnik, E.

    2009-01-01

    This paper examines if and how indoor climate systems are important for passive house certification. The research subjects are passive houses in Belgium, occupied by owner-clients. These have received a quality assurance certificate from an independent organization. Through interviews with the

  11. Programmable electronic safety systems

    International Nuclear Information System (INIS)

    Parry, R.R.

    1993-01-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement fail-safe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL

  12. Programmable Electronic Safety Systems

    International Nuclear Information System (INIS)

    Parry, R.

    1993-05-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement failsafe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL

  13. Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

    International Nuclear Information System (INIS)

    Sarkulov, M.K.

    2015-01-01

    There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

  14. Thermal fluid flow analysis in downcomer of JAERI passive safety light water reactor (JPSR)

    International Nuclear Information System (INIS)

    Kunii, K.; Iwamura, T.; Murao, Y.

    1995-01-01

    The residual heat for the JPSR (JAERI Passive Safety Light Water Reactor) is removed by a natural-circulation of coolant flowing through downcomer. The numerical analysis has been performed taking account of the downcomer being a three-dimensional annulus flow pass with the purposes to confirm the abilities of (1) approximation of three-dimensional thermal fluid flow in downcomer to simple one-dimensional one assumed on the preliminary design of the passive residual heat removal system and (2) achievement of an enough driving-force of the natural circulation to remove the residual heat. The following results were obtained : (1) Flow pattern in downcomer shows remarkable three-dimensionality (multi-dimensionality) at lower inlet flow rate not to be able to approximate to one-dimensional flow field. However, the temperature distribution does not deviate from uniform one so much even if the multi-dimensional flow such as large vortex arises. (2) It can be expected to obtain the required enough driving-force at a steady state in any case of inlet flow rate where multi-dimensional flow pattern appears. (3) The increase ratio of the driving-force with the time-integrated coolant amount can be estimated as two functional curves in case of higher and other lower inlet flow rates not dependent only on the respective inlet flow rate. (Author)

  15. On the status of the EFR Euro-Breeder and its passive safety features

    International Nuclear Information System (INIS)

    Marth, W.

    1992-01-01

    The Project of the EFR, the European Fast Reactor, is characterized by close European cooperation among power utilities, plant vendors, and research centers. In the present phase up until 1993 a consistent design of the nuclear part of the plant is being elaborated with the inclusion of a site-independent safety report. The most important design features, especially those in the field of passive safety, must be backed up by reliable R and D findings. These findings will enable the ad hoc Safety Club, a body of European safety experts, to pass its vote on the general licensability of the plant concept. (orig.) [de

  16. Passive heat removal system with injector-condenser

    Energy Technology Data Exchange (ETDEWEB)

    Soplenkov, K I [All-Russian Inst. of Nuclear Power Plant Operation, Electrogorsk Research and Engineering Centre of Nuclear Power Safety (Russian Federation)

    1996-12-01

    The system described in this paper is a passive system for decay heat removal from WWERs. It operates off the secondary side of the steam generators (SG). Steam is taken from the SG to operate a passive injector pump which causes secondary fluid to be pumped through a heat exchanger. Variants pass either water or steam from the SG through the heat exchanger. There is a passive initiation scheme. The programme for experimental and theoretical validation of the system is described. (author). 8 figs.

  17. Passivity Based Stabilization of Non-minimum Phase Nonlinear Systems

    Czech Academy of Sciences Publication Activity Database

    Travieso-Torres, J.C.; Duarte-Mermoud, M.A.; Zagalak, Petr

    2009-01-01

    Roč. 45, č. 3 (2009), s. 417-426 ISSN 0023-5954 R&D Projects: GA ČR(CZ) GA102/07/1596 Institutional research plan: CEZ:AV0Z10750506 Keywords : nonlinear systems * stabilisation * passivity * state feedback Subject RIV: BC - Control Systems Theory Impact factor: 0.445, year: 2009 http://library.utia.cas.cz/separaty/2009/AS/zagalak-passivity based stabilization of non-minimum phase nonlinear systems.pdf

  18. Optimization of Passive Coherent Receiver System Placement

    Science.gov (United States)

    2013-09-01

    spheroid object with a constant radar cross section (RCS). Additionally, the receiver and transmitters are assumed to be notional isotropic antennae...software- defined radio for equatorial plasma instability studies,” Radio Science, vol. 48, pp. 1–11. Aug. 2013. [2] P. C. Zhang and B. Y. Li, “Passive

  19. Problem of corium melt coolability in passive protection systems against severe accidents in the containment

    Directory of Open Access Journals (Sweden)

    Ali Kalvand

    2018-05-01

    Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.

  20. Design of passive decay heat removal system using thermosyphon for low temperature and low pressure pool type LWR

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; You, Byung Hyun; Jung, Yong Hun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In seawater desalination process which doesn't need high temperature steam, the reactor has profitability. KAIST has be developing the new reactor design, AHR400, for only desalination. For maximizing safety, the reactor requires passive decay heat removal system. In many nuclear reactors, DHR system is loop form. The DHR system can be designed simple by applying conventional thermosyphon, which is fully passive device, shows high heat transfer performance and simple structure. DHR system utilizes conventional thermosyphon and its heat transfer characteristics are analyzed for AHR400. For maximizing safety of the reactor, passive decay heat removal system are prepared. Thermosyphon is useful device for DHR system of low pressure and low temperature pool type reactor. Thermosyphon is operated fully passive and has simple structure. Bundle of thermosyphon get the goal to prohibit boiling in reactor and high pressure in reactor vessel.

  1. Systems Safety and Engineering Division

    Data.gov (United States)

    Federal Laboratory Consortium — Volpe's Systems Safety and Engineering Division conducts engineering, research, and analysis to improve transportation safety, capacity, and resiliency. We provide...

  2. ROADSIDE BARRIER AND PASSIVE SAFETY OF MOTORCYCLISTS ALONG EXCLUSIVE MOTORCYCLE LANES

    Directory of Open Access Journals (Sweden)

    A.B. IBITOYE

    2007-04-01

    Full Text Available The tremendous increase in number of motorcycles and fatalities in some ASEAN countries is becoming a main concern for the safety of motorcyclists along exclusive motorcycle lanes. The existing w-beam guardrail system along exclusive motorcycle lanes was originally designed to reduce severity of a crash when cars and trucks involve in run-off road accident – but not specifically to protect motorcyclists during such accident. However, the consequences of this guardrail design on the passive safety of motorcyclist have been given little consideration. Thus, Probability of the motorcyclists getting injured on collision with guardrail is higher compared to other motor vehicle’s driver. In order to investigate the passive safety of motorcyclists while in collision with this guardrail, this study carried out computer simulation of typical crash scenario and conducted a physical crash test to validate the simulation model. The study examines the crash mechanism as related to injury severity when motorcyclist interacts with W-beam guardrail. A three-dimensional computer simulation of a scaled Hybrid III 50th percentile Male dummy mounted on a motorcycle and colliding with W-beam guardrail was carried out. Multi-body model of motorcycle and finite element model of guardrail were developed with commercially available software called MADYMO. The simulation model is validated with a simple crash test conducted with same initial impact configuration. The subsequent simulations were set up for impacting the existing w-beam guardrail with 110 kg motorcycle using eighteen impact conditions that consist of impact angles 15o, 30o and 45o, impact speeds of 32, 48 and 60km/h as well as post spacing of 2m and 4m. The predicted rider’s injury risk criteria were used to assess safety of guardrail response to motorcyclists. The obtained results confirmed that the existing w-beam guardrail is not safe to motorcyclist, especially for the head injury at impact speed

  3. The CEA/DRN innovative R and D programme: significant studies on passive systems

    International Nuclear Information System (INIS)

    Fiorini, G.L.; Magistris, F. de; Dumaz, P.; Gautier, G.M.; Pignatel, J.F.; Richard, P.

    1999-01-01

    The work on passive systems is an essential item of the R and D programme for future reactors; it is structured following four main guidelines: Research, and validation of innovative solutions for the safety functions achievement; An enlarged assessment of the performances of passive systems; Extension of the data base and of the tools qualification range; Assessment of new plant operation modes. After a recalling on the general framework, the paper describes, following these guidelines, the status of the art of the main corresponding programmes within CEA/DRN. (author)

  4. Identification of passive shutdown system parameters in a metal fueled LMR

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1992-01-01

    This document discusses periodic testing of the passive shutdown system in a metal fueled liquid metal reactor which has been proposed as a Technical Specification requirement. In the approach to testing considered in this paper, perturbation experiments performed at normal operation are used to predict an envelope that bounds reactor response to flowrate, inlet temperature and external reactivity forcing functions. When the envelope for specific upsets lies within safety limits, one concludes that the passive shutdown system is operation properly for those upsets. Simulation results for the EBR-II reactor show that the response envelope for loss of flow and rod reactivity insertion events does indeed bound these events

  5. 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)

  6. Performance of ALMR passive decay heat removal system

    International Nuclear Information System (INIS)

    Boardman, C.E.; Hunsbedt, A.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the small (471 MWt) modular reactor to the environmental air by natural convection heat transfer. The system has no active components, requires no operator action to initiate, and is inherently reliable. The RVACS can perform its function under off-normal or degraded operating conditions without significant loss in performance. Several such events are described and the RVACS thermal performance for each is given and compared to the normal operation performance. The basic RVACS performance as well as the performance during several off-normal events have been updated to reflect design changes for recycled fuel with minor actinides for end of equilibrium cycle conditions. The performance results for several other off-normal events involving various degrees of RVACS air flow passage blockages are presented. The results demonstrated that the RVACS is unusually tolerant to a wide range of postulated faults. (author)

  7. Demonstration of passive safety features in EBR-II

    International Nuclear Information System (INIS)

    Planchon, H.P. Jr.; Golden, G.H.; Sackett, J.I.

    1987-01-01

    Two tests of great importance to the design of future commercial nuclear power plants were carried out in the Experimental Breeder Reactor-II on April 3, 1986. These tests, (viewed by about 60 visitors, including 13 foreign LMR specialists) were a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. In these tests, inherent feedback shut the reactor down without damage to the fuel or other reactor components. This resulted primarily from advantageous characteristics of the metal driver fuel used in EBR-II. Work is currently underway at EBR-II to develop a control strategy that promotes inherent safety characteristics, including survivability of transient overpower accidents. In parallel, work is underway at EBR-II on the development of state-of-the-art plant diagnostic techniques

  8. Reliability assessment of passive containment isolation system using APSRA methodology

    International Nuclear Information System (INIS)

    Nayak, A.K.; Jain, Vikas; Gartia, M.R.; Srivastava, A.; Prasad, Hari; Anthony, A.; Gaikwad, A.J.; Bhatia, S.; Sinha, R.K.

    2008-01-01

    In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) has been employed for evaluation of the reliability of passive systems. The methodology has been applied to the passive containment isolation system (PCIS) of the Indian advanced heavy water reactor (AHWR). In the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criterion. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the PCIS performance. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components is evaluated through a classical PSA treatment using the generic data. The reliability of the PCIS is evaluated from the probability of availability of the components for the success of the passive containment isolation system

  9. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    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.

  10. 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.

  11. Safety of huge systems

    International Nuclear Information System (INIS)

    Kondo, Jiro.

    1995-01-01

    Recently accompanying the development of engineering technology, huge systems tend to be constructed. The disaster countermeasures of huge cities become large problems as the concentration of population into cities is conspicuous. To make the expected value of loss small, the knowledge of reliability engineering is applied. In reliability engineering, even if a part of structures fails, the safety as a whole system must be ensured, therefore, the design having margin is carried out. The degree of margin is called redundancy. However, such design concept makes the structure of a system complex, and as the structure is complex, the possibility of causing human errors becomes high. At the time of huge system design, the concept of fail-safe is effective, but simple design must be kept in mind. The accident in Mihama No. 2 plant of Kansai Electric Power Co. and the accident in Chernobyl nuclear power station, and the accident of Boeing B737 airliner and the fatigue breakdown are described. The importance of safety culture was emphasized as the method of preventing human errors. Man-system interface and management system are discussed. (K.I.)

  12. Time-to-impact estimation in passive missile warning systems

    Science.gov (United States)

    Şahıngıl, Mehmet Cihan

    2017-05-01

    A missile warning system can detect the incoming missile threat(s) and automatically cue the other Electronic Attack (EA) systems in the suit, such as Directed Infrared Counter Measure (DIRCM) system and/or Counter Measure Dispensing System (CMDS). Most missile warning systems are currently based on passive sensor technology operating in either Solar Blind Ultraviolet (SBUV) or Midwave Infrared (MWIR) bands on which there is an intensive emission from the exhaust plume of the threatening missile. Although passive missile warning systems have some clear advantages over pulse-Doppler radar (PDR) based active missile warning systems, they show poorer performance in terms of time-to-impact (TTI) estimation which is critical for optimizing the countermeasures and also "passive kill assessment". In this paper, we consider this problem, namely, TTI estimation from passive measurements and present a TTI estimation scheme which can be used in passive missile warning systems. Our problem formulation is based on Extended Kalman Filter (EKF). The algorithm uses the area parameter of the threat plume which is derived from the used image frame.

  13. AC-600 passive containment cooling system performance research

    International Nuclear Information System (INIS)

    Jia Baoshan; Yu Jiyang; Shi Junying

    1997-01-01

    a code named PCCSAC which is able to predict both the evaporating film on the outside surface of the vessel and the condensed film on its inside is developed successfully. It is a special software tool to analyze the passive containment cooling system (PCCS) performance in the design of AC-600. The author includes the establishment of physical models, selection of numerical methods, debugging and verification of the code and application of the code in the AC-600 PCCS. In physical models, the fundamental conservation equations about various areas and heat conduction equations are established. In order to make the equations to meet the closed form of solution, a lot of structure formulae are complemented. After repeated selection and demonstration of the numerical methods, the backward difference method Gear which is generally used for stiff problem is chosen for the solution of ordinary differential equations derived from the physical models. The results of standard example calculated by the PCCSAC code and the COMMIX code which is used to analyze westinghouse AP-600 are same in the main. The reliability and validity are verified from the calculations. The PCCSAC code is applied in the calculations of two important LOCA used in the containment safety analyses. The sensitivity of main parameters in the system based on LOCA are studied. All the results are reasonable and in agreement with the theoretical analyses. It can be concluded that the PCCSAC code is able to be used for the analyses of AC-600 PCCS performance

  14. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  15. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  16. Concept of passive safe small reactor for distributed energy supply system

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

    2003-01-01

    This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

  17. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

  18. Passive safe small reactor for distributed energy supply system sited in water filled pit at seaside

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Imayoshi, Shou

    2003-01-01

    Japan Atomic Energy Research Institute has developed a Passive Safe Small Reactor for Distributed Energy Supply System (PSRD) concept. The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down device. For improvement of economy, simplification of the reactor system and long operation of the core over five years without refueling with low enriched UO 2 fuel rods are achieved. To avoid releasing the radioactive materials to the circumstance even if a hypothetical accident, the containment is submerged in a pit filled with seawater at a seaside. Refueling or maintenance of the reactor can be conducted using an exclusive barge instead of the reactor building. (author)

  19. Evaluation of the reliability of a passive system

    International Nuclear Information System (INIS)

    Bianchi, F.; Burgazzi, L.; D'Auria, F.; Galassi, G.M.; Ricotti, M.E.; Oriani, L.

    2001-01-01

    A passive system should be theoretically more reliable than an active one. In fact its operation is independent by any external input or energy and is relied only upon natural physical laws (e.g., gravity, natural circulation, etc.) and/or 'intelligent' use of the energy inherently available in the system (e.g., chemical reaction, decay heat, etc.). Nevertheless the passive system may fail its mission as consequences of component failures, deviation of physical phenomena, boundary and/or initial conditions from the expectation. This document describes at first the methodology developed by ENEA, in collaboration of University of Pisa and Polytechnic of Milano, allowing the evaluation of the reliability for a passive system, which operation is based on moving working fluids (type B and C, cf. IAEA). It reports the results of an exercise performed on a system, which operation is based on Natural Circulation.(author)

  20. 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.

  1. 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

  2. Passivity analysis and synthesis for uncertain time-delay systems

    Directory of Open Access Journals (Sweden)

    Magdi S. Mahmoud

    2001-01-01

    Full Text Available In this paper, we investigate the robust passivity analysis and synthesis problems for a class of uncertain time-delay systems. This class of systems arises in the modelling effort of studying water quality constituents in fresh stream. For the analysis problem, we derive a sufficient condition for which the uncertain time-delay system is robustly stable and strictly passive for all admissible uncertainties. The condition is given in terms of a linear matrix inequality. Both the delay-independent and delay-dependent cases are considered. For the synthesis problem, we propose an observer-based design method which guarantees that the closed-loop uncertain time-delay system is stable and strictly passive for all admissible uncertainties. Several examples are worked out to illustrate the developed theory.

  3. Accurately bearing measurement in non-cooperative passive location system

    International Nuclear Information System (INIS)

    Liu Zhiqiang; Ma Hongguang; Yang Lifeng

    2007-01-01

    The system of non-cooperative passive location based on array is proposed. In the system, target is detected by beamforming and Doppler matched filtering; and bearing is measured by a long-base-ling interferometer which is composed of long distance sub-arrays. For the interferometer with long-base-line, the bearing is measured accurately but ambiguously. To realize unambiguous accurately bearing measurement, beam width and multiple constraint adoptive beamforming technique is used to resolve azimuth ambiguous. Theory and simulation result shows this method is effective to realize accurately bearing measurement in no-cooperate passive location system. (authors)

  4. The PANDA tests for the SWR 1000 passive containment cooling system

    International Nuclear Information System (INIS)

    Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

    1999-01-01

    Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R+D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR 1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

  5. Study on development of active-passive rehabilitation system for upper limbs: Hybrid-PLEMO

    International Nuclear Information System (INIS)

    Kikuchi, T; Jin, Y; Fukushima, K; Akai, H; Furusho, J

    2009-01-01

    In recent years, many researchers have studied the potential of using robotics technology to assist and quantify the motor functions for neuron-rehabilitation. Some kinds of haptic devices have been developed and evaluated its efficiency with clinical tests, for example, upper limb training for patients with spasticity after stroke. Active-type (motor-driven) haptic devices can realize a lot of varieties of haptics. But they basically require high-cost safety system. On the other hand, passive-type (brake-based) haptic devices have inherent safety. However, the passive robot system has strong limitation on varieties of haptics. There are not sufficient evidences to clarify how the passive/active haptics effect to the rehabilitation of motor skills. In this paper, we developed an active-passive-switchable rehabilitation system with ER clutch/brake device named 'Hybrid-PLEMO' in order to address these problems. In this paper, basic structures and haptic control methods of the Hybrid-PLEMO are described.

  6. Study on development of active-passive rehabilitation system for upper limbs: Hybrid-PLEMO

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, T; Jin, Y; Fukushima, K; Akai, H; Furusho, J [Department of Mechanical Engineering, Graduate School of Engineering, Osaka University, Osaka (Japan)], E-mail: kikuchi@mech.eng.osaka-u.ac.jp

    2009-02-01

    In recent years, many researchers have studied the potential of using robotics technology to assist and quantify the motor functions for neuron-rehabilitation. Some kinds of haptic devices have been developed and evaluated its efficiency with clinical tests, for example, upper limb training for patients with spasticity after stroke. Active-type (motor-driven) haptic devices can realize a lot of varieties of haptics. But they basically require high-cost safety system. On the other hand, passive-type (brake-based) haptic devices have inherent safety. However, the passive robot system has strong limitation on varieties of haptics. There are not sufficient evidences to clarify how the passive/active haptics effect to the rehabilitation of motor skills. In this paper, we developed an active-passive-switchable rehabilitation system with ER clutch/brake device named 'Hybrid-PLEMO' in order to address these problems. In this paper, basic structures and haptic control methods of the Hybrid-PLEMO are described.

  7. The PANDA tests for the SWR 1000 passive containment cooling system

    International Nuclear Information System (INIS)

    Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Meseth, J.; Yadigaroglu, G.

    1999-01-01

    Since 1992, Siemens has been developing the SWR 1000, a new boiling water reactor with passive safety features. This development has been performed in close co-operation with the German nuclear utilities and with support from various European partners. Within the European Union sponsored project 'BWR R and D Cluster for Innovative Passive Safety Systems' and a bilateral contract between Siemens and the Paul Scherrer Institute, the passive containment cooling system of the SWR 1000 design has been investigated in the large-scale PANDA test facility at the Paul Scherrer Institute. A series of six tests were performed to simulate transients selected to cover a range of failure assumptions and accident severity, including core heat up and hydrogen generation. The results graphically demonstrate the self regulating character of the passive heat removal systems and their effectiveness, even under severe load, in limiting the containment pressurisation. Some tentative conclusions for the SWR1000 are drawn, to be established by detailed analyses of the data, to support models and codes for application to plant transients. (author)

  8. Utilizing the Fast Flux Test Facility for international passive safety testing

    International Nuclear Information System (INIS)

    Shen, P.K.; Padilla, A.; Lucoff, D.M.; Waltar, A.E.

    1991-01-01

    A two-phased approach has been undertaken in the Fast Flux Test Facility (FFTF) to conduct passive safety testing. Phase I (1986 to 1987) was structured to obtain an initial understanding of the reactivity feedback components. The planned Phase II (1992 to 1993) international program will extend the testing to include static and dynamic feedback measurements, transient and demonstration tests, and gas expansion module (GEM) reactivity tests. The primary objective is to meet the needs for safety analysis code validation, with particular emphasis on reducing the uncertainties associated with structure reactivity feedback. Program scope and predicted FFTF responses are discussed and illustrated. (author)

  9. 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

  10. Multiple-Active Multiple-Passive Antenna Systems and Applications

    DEFF Research Database (Denmark)

    Tsakalaki, Elpiniki

    2013-01-01

    -passive (MAMP) antenna topologies, as explained in Sect. 8.1. Then, Sect. 8.2 proposes MAMP antenna structures with application to reconfigurable MIMO transmission in the presence of antenna mutual coupling under poor scattering channel conditions. For this purpose, the section presents an adaptive MAMP antenna...... system capable of changing its transmission parameters via passive radiators attached to tunable loads, according to the structure of the RF propagation channel. The hybrid MAMP array structure can be tractably analyzed using the active element response vector (instead of the classical steering vector...... adaptive MAMP system can be limited to practical dimensions whereas the passive antennas require no extra RF hardware, thus meeting the cost, space, and power constrains of the users’ mobile terminals. The simulation results show that the adaptive MAMP system, thanks to its “adaptivity”, is able to achieve...

  11. [Post-licensure passive safety surveillance of rotavirus vaccines: reporting sensitivity for intussusception].

    Science.gov (United States)

    Pérez-Vilar, S; Díez-Domingo, J; Gomar-Fayos, J; Pastor-Villalba, E; Sastre-Cantón, M; Puig-Barberà, J

    2014-08-01

    The aims of this study were to describe the reports of suspected adverse events due to rotavirus vaccines, and assess the reporting sensitivity for intussusception. Descriptive study performed using the reports of suspected adverse events following rotavirus vaccination in infants aged less than 10 months, as registered in the Pharmacovigilance Centre of the Valencian Community during 2007-2011. The reporting rate for intussusception was compared to the intussusception rate in vaccinated infants obtained using the hospital discharge database (CMBD), and the regional vaccine registry. The adverse event reporting rate was 20 per 100,000 administered doses, with the majority (74%) of the reports being classified as non-serious. Fever, vomiting, and diarrhea were the adverse events reported more frequently. Two intussusception cases, which occurred within the first seven days post-vaccination, were reported as temporarily associated to vaccination. The reporting sensitivity for intussusception at the Pharmacovigilance Centre in the 1-7 day interval following rotavirus vaccination was 50%. Our results suggest that rotavirus vaccines have, in general, a good safety profile. Intussusception reporting to the Pharmacovigilance Centre shows sensitivity similar to other passive surveillance systems. The intussusception risk should be further investigated using well-designed epidemiological studies, and evaluated in comparison with the well-known benefits provided by these vaccines. Copyright © 2013 Asociación Española de Pediatría. Published by Elsevier Espana. All rights reserved.

  12. Safety significance evaluation system

    International Nuclear Information System (INIS)

    Lew, B.S.; Yee, D.; Brewer, W.K.; Quattro, P.J.; Kirby, K.D.

    1991-01-01

    This paper reports that the Pacific Gas and Electric Company (PG and E), in cooperation with ABZ, Incorporated and Science Applications International Corporation (SAIC), investigated the use of artificial intelligence-based programming techniques to assist utility personnel in regulatory compliance problems. The result of this investigation is that artificial intelligence-based programming techniques can successfully be applied to this problem. To demonstrate this, a general methodology was developed and several prototype systems based on this methodology were developed. The prototypes address U.S. Nuclear Regulatory Commission (NRC) event reportability requirements, technical specification compliance based on plant equipment status, and quality assurance assistance. This collection of prototype modules is named the safety significance evaluation system

  13. Blind spot detection & passive lane change assist systems

    NARCIS (Netherlands)

    Surovtcev, I.

    2015-01-01

    The project goal was design and implementation of proof-of-concept for two systems that aim to tackle the blind spot problem of for the commercial vehicles: Blind Spot Detection and Passive Lane Change Assist functions. The system implementation was done using Rapid Control Prototype (RCP) hardware.

  14. Analysis for thermal fluid dynamics in downcomer of JAERI passive safety reactor (JPSR)

    International Nuclear Information System (INIS)

    Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

    1995-01-01

    The driving-force of the natural circulation in the residual heat removal system for the JPSR (JAERI Passive Safety Reactor) under a steady condition is given as a gravity force based on the density (temperature) difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The downcomer is a very important flow pass in the system to obtain the enough driving-force because the flow pass has a three-dimensional annulus geometry long in vertical and circumference directions respectively and narrow in radius direction so that the thermal fluid flow pattern in downcomer directly relates to generation of the density difference. The density difference could naturally become smaller unless the coolant flowing into downcomer spreads widely in the whole region of it. The numerical analysis has been performed taking account of the downcomer being a three-dimensional annulus flow pass with the purposes to investigate the possibilities of the followings: (1) promotion of making the flow pattern and temperature distribution uniform in downcomer by applying a mechanical device at the inlet part of downcomer (installing a baffle) to increase the driving-force of the natural circulation, (2) achievement of an enough driving-force of the natural circulation to remove the residual heat, (3) approximation of three-dimensional thermal fluid flow in downcomer to simple one-dimensional one assumed on the preliminary design of the passive residual heat removal system. The following conclusions were obtained: (1) The effect of the baffle on the driving-force of natural circulation is little being considered due to the enhancing of mixing on thermal fluid flow in case with baffle, (2) Though the flow pattern becomes three-dimensional in some case such as large vortex flow not to be able to approximate simply to one-dimensional, the required driving-force can be obtained, (3) The driving-force can be estimated as the almost same functional value for time

  15. Evaluating safety management system implementation

    International Nuclear Information System (INIS)

    Preuss, M.

    2009-01-01

    Canada is committed to not only maintaining, but also improving upon our record of having one of the safest aviation systems in the world. The development, implementation and maintenance of safety management systems is a significant step towards improving safety performance. Canada is considered a world leader in this area and we are fully engaged in implementation. By integrating risk management systems and business practices, the aviation industry stands to gain better safety performance with less regulatory intervention. These are important steps towards improving safety and enhancing the public's confidence in the safety of Canada's aviation system. (author)

  16. Passive ALWR safety: the ALPHA project at Switzerland's PSI - a progress report

    International Nuclear Information System (INIS)

    Yadigaroglu, G.; Varadi, G.; Dreier, J.

    1995-01-01

    The Paul Scherrer Institute (PSI) initiated the ALPHA project in 1991 for the experimental and analytical investigation of the long-term decay heat removal from the containment of the next generation of 'passive' advanced light water reactors (ALWRs). The dynamic containment response to such systems, as well as containment phenomena, are investigated. The ALPHA project includes integral system tests in the large-scale (1:25 in volume) PANDA facility; the smaller-scale separate effects LINX series of tests related to various passive containment mixing, stratification, and condensation phenomena in the presence of non-condensable gases; the AIDA tests on the behavior of aerosols in passive containment cooling systems (PCCS); and supporting analytical work. The project has been, so far, directed mainly to the investigation of the General Electric simplified boiling water reactor (SBWR) PCCS and related phenomena. (author) 2 figs., 4 refs

  17. Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

    International Nuclear Information System (INIS)

    Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

    2014-01-01

    To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

  18. Uncertainty analysis methods for estimation of reliability of passive system of VHTR

    International Nuclear Information System (INIS)

    Han, S.J.

    2012-01-01

    An estimation of reliability of passive system for the probabilistic safety assessment (PSA) of a very high temperature reactor (VHTR) is under development in Korea. The essential approach of this estimation is to measure the uncertainty of the system performance under a specific accident condition. The uncertainty propagation approach according to the simulation of phenomenological models (computer codes) is adopted as a typical method to estimate the uncertainty for this purpose. This presentation introduced the uncertainty propagation and discussed the related issues focusing on the propagation object and its surrogates. To achieve a sufficient level of depth of uncertainty results, the applicability of the propagation should be carefully reviewed. For an example study, Latin-hypercube sampling (LHS) method as a direct propagation was tested for a specific accident sequence of VHTR. The reactor cavity cooling system (RCCS) developed by KAERI was considered for this example study. This is an air-cooled type passive system that has no active components for its operation. The accident sequence is a low pressure conduction cooling (LPCC) accident that is considered as a design basis accident for the safety design of VHTR. This sequence is due to a large failure of the pressure boundary of the reactor system such as a guillotine break of coolant pipe lines. The presentation discussed the obtained insights (benefit and weakness) to apply an estimation of reliability of passive system

  19. The aviation safety reporting system

    Science.gov (United States)

    Reynard, W. D.

    1984-01-01

    The aviation safety reporting system, an accident reporting system, is presented. The system identifies deficiencies and discrepancies and the data it provides are used for long term identification of problems. Data for planning and policy making are provided. The system offers training in safety education to pilots. Data and information are drawn from the available data bases.

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

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

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

  1. 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.

  2. Appraisal of Passive and Active Fire Protection Systems in Student’s Accommodation

    Directory of Open Access Journals (Sweden)

    Ismail I.

    2014-03-01

    Full Text Available Fire protection systems are very important systems that must be included in buildings. They have a great significance in reducing or preventing the occurrences of fire. This paper presents an assessment of fire protection systems in student’s accommodation. Student accommodation is a particular type of building that provides shelter for students at University. In addition, it is also supposed to be an attractive environment, conducive to learning, and importantly, safe for occupation. The fire safety of occupants in a building, must be in accordance with the requirements of the building’s code. Therefore, the design of the building must comply with the Uniform Building By-Law (UBBL 1984 of Malaysia, and provide all of the required safety features. This paper describes the findings from investigations of passive and active fire protection systems installed in buildings, based on fire safety requirements, UBBL (1984.

  3. Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

    Energy Technology Data Exchange (ETDEWEB)

    Vierow, Karen

    2005-08-29

    This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

  4. evaluation of a modified passive solar housing system for poultry

    African Journals Online (AJOL)

    User

    The hourly efficiency of the solar brick passive system was estimated at about 78.42% in a day of May and ... to high cost and unavailability of kerosene in most developing .... sulted in intermittent rainfall, cloud cover and sunshine. From the ...

  5. Development on design methodology of PWR passive containment system

    International Nuclear Information System (INIS)

    Lee, Seong Wook

    1998-02-01

    The containment is the most important barrier against the release of radioactive materials into the environment during accident conditions of nuclear power plants. Therefore the development of a reliable containment cooling system is one of key areas in advanced reactor development. To enhance the safety of the containment system, many new containment system designs have been proposed and developed in the world. Several passive containment cooling system (PCCS) concepts for both steel and concrete containment systems are overviewed and assessed comparatively. Major concepts considered are: (a) the spray of water on the outer surface of a steel containment from an elevated tank, (b) an external moat for a steel containment, (c) a suppression pool for a concrete containment, and (d) combination of the internal spray and internal or external condensers for a concrete containment. Emphasis is given to the heat removal principles, the required heat transfer area, system complexity and operational reliability. As one of conceptual design steps of containment, a methodology based on scaling principles is proposed to determine the containment size according to the power level. The AP600 containment system is selected as the reference containment to which the scaling laws are applied. Governing equations of containment pressure are set up in consideration of containment behavior in accident conditions. Then, the dimensionless numbers, which characterize the containment phenomena, are derived for the blowdown dominant and decay heat dominant stage, respectively. The important phenomena in blowdown stage are mass and energy sources and their absorption in containment atmosphere or containment structure, while heat transfer to the outer environment becomes important in decay heat stage. Based on their similarity between the prototype and the model, the containment sizes are determined for higher power levels and are compared with the SPWR containment design values available

  6. Safety logic systems of PFBR

    International Nuclear Information System (INIS)

    Sambasivan, S. Ilango

    2004-01-01

    Full text : PFBR is provided with two independent, fast acting and diverse shutdown systems to detect any abnormalities and to initiate safety action. Each system consists of sensors, signal processing systems, logics, drive mechanisms and absorber rods. The absorber rods of the first system are Control and Safety Rods (CSR) and that of the second are called as Diverse Safety Rods (DSR). There are nine CSR and three DSR. While CSR are used for startup, control of reactor power, controlled shutdown and SCRAM, the DSR are used only for SCRAM. The respective drive mechanisms are called as CSRDM and DSRDM. Each of these two systems is capable of executing the shutdown satisfactorily with single failure criteria. Two independent safety logic systems based on diverse principles have been designed for the two shut down systems. The analog outputs of the sensors of Core Monitoring Systems comprising of reactor flux monitoring, core temperature monitoring, failed fuel detection and core flow monitoring systems are processed and converted into binary signals depending on their instantaneous values. Safety logic systems receive the binary signals from these core-monitoring systems and process them logically to protect the reactor against postulated initiating events. Neutronic and power to flow (P/Q) signals form the inputs to safety logic system-I and temperature signals are inputs to the safety logic system II. Failed fuel detection signals are processed by both the shut down systems. The two logic systems to actuate the safety rods are also based on two diverse designs and implemented with solid-state devices to meet all the requirements of safety systems. Safety logic system I that caters to neutronic and P/Q signals is designed around combinational logic and has an on-line test facility to detect struck at faults. The second logic system is based on dynamic logic and hence is inherently safe. This paper gives an overview of the two logic systems that have been

  7. Power system harmonics and passive filter designs

    CERN Document Server

    Das, J C

    2015-01-01

    J.C. Das is a consultant of electrical power systems at Power Systems Studies, Inc., USA. He is Life Fellow of IEEE (UK), Fellow of IET (India), and has authored approximately sixty technical papers and published 190 study reports of real-world power systems. He is the author of three books including ARC Flash Hazard Analysis and Mitigation. He is a registered P.E. in the states of Georgia and Oklahoma, C.Eng. in UK, and Eur Ing in Europe. J. C. Das is also a member of CIGRE, Federation of European Engineers, and other technical associations and organizations.

  8. Aviation Safety Hotline Information System -

    Data.gov (United States)

    Department of Transportation — The Aviation Safety Hotline Information System (ASHIS) collects, stores, and retrieves reports submitted by pilots, mechanics, cabin crew, passengers, or the public...

  9. Traceability of Software Safety Requirements in Legacy Safety Critical Systems

    Science.gov (United States)

    Hill, Janice L.

    2007-01-01

    How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

  10. Numerical Analysis of a Passive Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Kim, Taejoon; Ha, Huiun; Heo, Sun

    2014-01-01

    The passive Containment Filtered Venting system (CFVS) does not have principally any kind of isolation valves or filtering devices which need periodic maintenance. In this study, the hydro-thermal analysis is presented to investigate the existence of flow instability in the passive CFVS and its performance under the pressure change of APR+ containment building with LB-LOCA M/E data. The Passive Containment Filtered Venting System was suggested as a part in i-Power development project and the operation mechanism was investigated by numerical modeling and simulation using GOTHIC8.0 system code. There are four Phases for consideration to investigate the pressurization of the containment building, loss of hydrostatic head in the pipe line of CFVS, opening of pipe line and gas ejection to the coolant tank, and the head recovery inside the pipe as the containment gas exhausted. The simulation results show that gas generation rate determine the timing of head recovery in the CFVS pipe line and that the equipment of various devices inducing pressure loss at the pipe can give the capacity of Phase control of the passive CFVS operation

  11. Power Control for Passive QAM Multisensor Backscatter Communication Systems

    Directory of Open Access Journals (Sweden)

    Shengbo Hu

    2017-01-01

    Full Text Available To achieve good quality of service level such as throughput, power control is of great importance to passive quadrature amplitude modulation (QAM multisensor backscatter communication systems. First, we established the RF energy harvesting model and gave the energy condition. In order to minimize the interference of subcarriers and increase the spectral efficiency, then, the colocated passive QAM backscatter communication signal model is presented and the nonlinear optimization problems of power control are solved for passive QAM backscatter communication systems. Solutions include maximum and minimum access interval, the maximum and minimum duty cycle, and the minimal RF-harvested energy under the energy condition for node operating. Using the solutions above, the maximum throughput of passive QAM backscatter communication systems is analyzed and numerical calculation is made finally. Numerical calculation shows that the maximal throughput decreases with the consumed power and the number of sensors, and the maximum throughput is decreased quickly with the increase of the number of sensors. Especially, for a given consumed power of sensor, it can be seen that the throughput decreases with the duty cycle and the number of sensors has little effect on the throughput.

  12. Design and analysis of a new passive residual heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Xing [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Peng, Minjun, E-mail: heupmj@163.com [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China); Yuan, Xiao [Guangxi Fangchenggang Nuclear Power Co., Ltd (China); Xia, Genglei [Key Subject Laboratory of Nuclear Safety and Simulation Technology, Harbin Engineering University, Harbin, Heilongjiang 150001 (China)

    2016-07-15

    Highlights: • An air cooling passive residual heat removal System (PRHRs) is designed. • Using RELAP5/MOD3.4 code to analyze the operation characteristics of the PRHRs. • Noncondensable gas is used to simulate the hydrodynamic behavior in the air cooling tower. • The natural circulations could respectively establish in the primary circuit and the PRHRs circuit. • The PRHRs could remove the residual heat effectively. - Abstract: The inherent safety functions will mitigate the consequences of the accidents, and it can be accomplished through the passive safety systems which employed in the typical pressurized water reactor (PWR). In this paper, a new passive residual heat removal system (PRHRS) is designed for a typical nuclear power plant. PRHRS consists of a steam generator (SG), a cooling tank with two groups of cooling pipes, an air-cooling heat exchanger (AHX), an air-cooling tower, corresponding pipes and valves. The cooling tank which works as an intermediate buffer device is used to transfer the core decay heat to the AHX, and then the core decay heat will be removed to the atmosphere finally. The RELAP5/MOD3.4 code is used to analyze the operation characteristics of PRHRS and the primary loop system. It shows PRHRS could remove the decay heat from the primary loop effectively, and the natural circulations can be established in the primary circuit and the PRHRS circuit respectively. Furthermore, the sensitivity study has also been done to research the effect of various factors on the heat removal capacity.

  13. London 2012 Paralympic swimming: passive drag and the classification system.

    Science.gov (United States)

    Oh, Yim-Taek; Burkett, Brendan; Osborough, Conor; Formosa, Danielle; Payton, Carl

    2013-09-01

    The key difference between the Olympic and Paralympic Games is the use of classification systems within Paralympic sports to provide a fair competition for athletes with a range of physical disabilities. In 2009, the International Paralympic Committee mandated the development of new, evidence-based classification systems. This study aims to assess objectively the swimming classification system by determining the relationship between passive drag and level of swimming-specific impairment, as defined by the current swimming class. Data were collected on participants at the London 2012 Paralympic Games. The passive drag force of 113 swimmers (classes 3-14) was measured using an electro-mechanical towing device and load cell. Swimmers were towed on the surface of a swimming pool at 1.5 m/s while holding their most streamlined position. Passive drag ranged from 24.9 to 82.8 N; the normalised drag (drag/mass) ranged from 0.45 to 1.86 N/kg. Significant negative associations were found between drag and the swimming class (τ = -0.41, p < 0.01) and normalised drag and the swimming class (τ = -0.60, p < 0.01). The mean difference in drag between adjacent classes was inconsistent, ranging from 0 N (6 vs 7) to 11.9 N (5 vs 6). Reciprocal Ponderal Index (a measure of slenderness) correlated moderately with normalised drag (r(P) = -0.40, p < 0.01). Although swimmers with the lowest swimming class experienced the highest passive drag and vice versa, the inconsistent difference in mean passive drag between adjacent classes indicates that the current classification system does not always differentiate clearly between swimming groups.

  14. 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)

  15. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

    1995-01-01

    The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

  16. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    Naito, Morimasa

    2006-01-01

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

  17. 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

  18. Immunizations: Active vs. Passive

    Science.gov (United States)

    ... Issues Health Issues Health Issues Conditions Injuries & Emergencies Vaccine Preventable Diseases ... Children > Safety & Prevention > Immunizations > Immunizations: Active vs. Passive Safety & ...

  19. Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI Line Break

    Directory of Open Access Journals (Sweden)

    A.S. Ekariansyah

    2016-08-01

    Full Text Available Generation II Nuclear Power Plants (NPPs have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

  20. System safety engineering analysis handbook

    Science.gov (United States)

    Ijams, T. E.

    1972-01-01

    The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

  1. Innovation in the Safety of nuclear systems: fundamental aspects

    International Nuclear Information System (INIS)

    Herranz, L. E.

    2009-01-01

    Safety commercial nuclear reactors has been an indispensable condition for future enlargement of power generation based on nuclear technology. Its fundamental principle, defence in depth, far from being outdated, is still adopted as a key foundation in the advanced nuclear system (generations III and IV). Nevertheless, the cumulative experience gained in the operation and maintenance of nuclear reactors, the development of methodologies like the probabilistic safety analysis, the use of passive safety systems and, even, the inherent characteristics of some new design (which exclude accident scenarios), allow estimating safety figures of merit even more outstanding that those achieved in the second generation of nuclear reactors. This safety innovation of upcoming nuclear reactors has entailed a huge investigation program (generation III) that will be focused on optimizing and demonstrating the postulated safety of future nuclear systems (Generation IV). (Author)

  2. The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

    International Nuclear Information System (INIS)

    Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

    1987-03-01

    A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

  3. Evaluation of conceptual Heat Exchanger Design for passive containment cooling system of SMART

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Ki; Hong, Soon Joon [FNC Tech., Yongin (Korea, Republic of); Kim, Young In; Kim, Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    PCCS(Passive containment cooling system) is the passive safety system which ultimately removes the reactor decay heat. Cooling performance of the air-cooled type and water-circulation cooling type of PCCS were analyzed using CAP version 2.21. The analysis results show the water-circulation cooling PCCS is more effective in lowering the peak pressure and temperature in the containment building. However, the air-cooled PCCS is more effective to the long-term cooling. From this study, the efficiency evaluation results for the two PCCS designs are obtained. These results may be applied in the PCCS design improvement. Moreover, these results will be used as a reference for the later PCCS design and analysis.

  4. Assessment of the effect of nitrogen gas on passive containment cooling system performance

    International Nuclear Information System (INIS)

    Ha, Huiun; Suh, Jungsoo

    2016-01-01

    As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR

  5. Assessment of the effect of nitrogen gas on passive containment cooling system performance

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Suh, Jungsoo [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR.

  6. Antenna array geometry optimization for a passive coherent localisation system

    Science.gov (United States)

    Knott, Peter; Kuschel, Heiner; O'Hagan, Daniel

    2012-11-01

    Passive Coherent Localisation (PCL), also known as Passive Radar, making use of RF sources of opportunity such as Radio or TV Broadcasting Stations, Cellular Phone Network Base Stations, etc. is an advancing technology for covert operation because no active radar transmitter is required. It is also an attractive addition to existing active radar stations because it has the potential to discover low-flying and low-observable targets. The CORA (Covert Radar) experimental passive radar system currently developed at Fraunhofer-FHR features a multi-channel digital radar receiver and a circular antenna array with separate elements for the VHF- and the UHF-range and is used to exploit alternatively Digital Audio (DAB) or Video Broadcasting (DVB-T) signals. For an extension of the system, a wideband antenna array is being designed for which a new discone antenna element has been developed covering the full DVB-T frequency range. The present paper describes the outline of the system and the numerical modelling and optimisation methods applied to solve the complex task of antenna array design: Electromagnetic full wave analysis is required for the parametric design of the antenna elements while combinatorial optimization methods are applied to find the best array positions and excitation coefficients for a regular omni-directional antenna performance. The different steps are combined in an iterative loop until the optimum array layout is found. Simulation and experimental results for the current system will be shown.

  7. A new variable stiffness suspension system: passive case

    Directory of Open Access Journals (Sweden)

    O. M. Anubi

    2013-02-01

    Full Text Available This paper presents the design, analysis, and experimental validation of the passive case of a variable stiffness suspension system. The central concept is based on a recently designed variable stiffness mechanism. It consists of a horizontal control strut and a vertical strut. The main idea is to vary the load transfer ratio by moving the location of the point of attachment of the vertical strut to the car body. This movement is controlled passively using the horizontal strut. The system is analyzed using an L2-gain analysis based on the concept of energy dissipation. The analyses, simulation, and experimental results show that the variable stiffness suspension achieves better performance than the constant stiffness counterpart. The performance criteria used are; ride comfort, characterized by the car body acceleration, suspension deflection, and road holding, characterized by tire deflection.

  8. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

    Delage, M.; Giroux, C.

    1990-12-01

    Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

  9. System safety education focused on system management

    Science.gov (United States)

    Grose, V. L.

    1971-01-01

    System safety is defined and characteristics of the system are outlined. Some of the principle characteristics include role of humans in hazard analysis, clear language for input and output, system interdependence, self containment, and parallel analysis of elements.

  10. Steady state flow evaluations for passive auxiliary feedwater system of APR

    International Nuclear Information System (INIS)

    Park, Jongha; Kim, Jaeyul; Seong, Hoje; Kang, Kyoungho

    2012-01-01

    This paper briefly introduces a methodology to evaluate steady state flow of APR+ Passive Auxiliary Feedwater System (PAFS). The PAFS is being developed as a safety grade passive system to completely replace the existing active Auxiliary Feedwater System (AFWS). Natural circulation cooling can be generally classified into the single-phase, two-phase, and boiling-condensation modes. The PAF is designed to be operated in a boiling-condensation natural circulation mode. The steady-state flow rate should be equal to the steady-state boiling/condensation rate determined by the steady-state energy and momentum balances in the PAFS. The determined steady-state flow rate can be used in the design optimization for the natural circulation loop of the PAFS through the steady-state momentum balance. Since the retarding force, which is to be balanced by the driving force in the natural circulation system design depends on the reliable evaluation of the success of a natural circulation system design depends on the reliable evaluation of the pressure loss coefficients. In PAFS, the core decay heat is released by natural circulation flow between the S G secondary side and the Passive Condensation Heat Exchanger (PCHX) that is immersed in the Passive Condensation Cooling Tank (PCCT). The PCCT is located on the top of Auxiliary building The driving force is determined by the difference between the S/G (heat Source) secondary water level and condensation liquid (heat sink) level. It will overcome retarding force at flowrate in the system, which is determined by vaporization and condensation of the steam which is generated at the S/G by the latent heat in system. In this study, the theoretical method to estimate the steady state flow rate in boiling-condensation natural circulation system is developed and compared with test results

  11. Software Safety Risk in Legacy Safety-Critical Computer Systems

    Science.gov (United States)

    Hill, Janice L.; Baggs, Rhoda

    2007-01-01

    Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

  12. LWIR passive perception system for stealthy unmanned ground vehicle night operations

    Science.gov (United States)

    Lee, Daren; Rankin, Arturo; Huertas, Andres; Nash, Jeremy; Ahuja, Gaurav; Matthies, Larry

    2016-05-01

    Resupplying forward-deployed units in rugged terrain in the presence of hostile forces creates a high threat to manned air and ground vehicles. An autonomous unmanned ground vehicle (UGV) capable of navigating stealthily at night in off-road and on-road terrain could significantly increase the safety and success rate of such resupply missions for warfighters. Passive night-time perception of terrain and obstacle features is a vital requirement for such missions. As part of the ONR 30 Autonomy Team, the Jet Propulsion Laboratory developed a passive, low-cost night-time perception system under the ONR Expeditionary Maneuver Warfare and Combating Terrorism Applied Research program. Using a stereo pair of forward looking LWIR uncooled microbolometer cameras, the perception system generates disparity maps using a local window-based stereo correlator to achieve real-time performance while maintaining low power consumption. To overcome the lower signal-to-noise ratio and spatial resolution of LWIR thermal imaging technologies, a series of pre-filters were applied to the input images to increase the image contrast and stereo correlator enhancements were applied to increase the disparity density. To overcome false positives generated by mixed pixels, noisy disparities from repeated textures, and uncertainty in far range measurements, a series of consistency, multi-resolution, and temporal based post-filters were employed to improve the fidelity of the output range measurements. The stereo processing leverages multi-core processors and runs under the Robot Operating System (ROS). The night-time passive perception system was tested and evaluated on fully autonomous testbed ground vehicles at SPAWAR Systems Center Pacific (SSC Pacific) and Marine Corps Base Camp Pendleton, California. This paper describes the challenges, techniques, and experimental results of developing a passive, low-cost perception system for night-time autonomous navigation.

  13. 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)

  14. A passive automated personnel accountability system for reactor emergency preparedness

    International Nuclear Information System (INIS)

    Zimmerman, R.O.; DeLisle, G.V.; Hickey, E.E.

    1988-04-01

    In 1985 a project was undertaken at the N Reactor on the Hanford Site to develop an automated personnel accountability system to ensure accountability of all personnel within 30 minutes of a site evacuation. The decision to develop such a system was made after a full-scale evacuation drill showed that the manual accountability system in use at the time was inadequate to meet the 30-minute requirement. Accountability systems at commercial nuclear power plants were evaluated, but found to be unsuitable because they were not passive, that is, they required action on part of the user for the system to work. Approximately 2500 people could be required to evacuate the 100-N Area. Therefore, a card key system or badge exchange system was judged not to be feasible. A passive accountability system was desired for N Reactor to allow personnel to enter and leave the site in a more timely manner. To meet the need for an automated accountability system at N Reactor, a special Evacuation Accountability System (EVACS) was designed and developed. The EVACS system has three basic components: the transponder, a credit card-sized device worn with the security badge; portal monitors, which are electronically activated by the transponder; and a computer information system that contains the personnel data base. Each person wearing a transponder is accounted for automatically by walking through a portal. In this paper, a description of the hardware and software will be presented, together with problems encountered and lessons learned while adapting an existing technology to this particular use. The system is currently installed and requires acceptance testing before becoming operational

  15. 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

  16. A completely passive continuous flow solar water purification system

    Energy Technology Data Exchange (ETDEWEB)

    Duff, William S.; Hodgson, David A. [Dept. of Mechanical Enginnering, Colorado State Univ., Fort Collins, CO (United States)

    2008-07-01

    Water-borne pathogens in developing countries cause several billion cases of disease and up to 10 million deaths each year, at least half of which are children. Solar water pasteurization is a potentially cost-effective, robust and reliable solution to these problems. A completely passively controlled solar water pasteurization system with a total collector area of 0.45 m{sup 2} has been constructed. The system most recently tested produced 337 litres per m{sup 2} of collector area of treated water on a sunny day. We developed our completely passive density-driven solar water pasteurization system over a five year span so that it now achieves reliable control for all possible variations in solar conditions. We have also substantially increased its daily pure water production efficiency over the same period. We will discuss the performance of our water purification system and provide an analyses that demonstrates that the system insures safe purified water production at all times. (orig.)

  17. Inter comparison of REPAS and APSRA methodologies for passive system reliability analysis

    International Nuclear Information System (INIS)

    Solanki, R.B.; Krishnamurthy, P.R.; Singh, Suneet; Varde, P.V.; Verma, A.K.

    2014-01-01

    The increasing use of passive systems in the innovative nuclear reactors puts demand on the estimation of the reliability assessment of these passive systems. The passive systems operate on the driving forces such as natural circulation, gravity, internal stored energy etc. which are moderately weaker than that of active components. Hence, phenomenological failures (virtual components) are equally important as that of equipment failures (real components) in the evaluation of passive systems reliability. The contribution of the mechanical components to the passive system reliability can be evaluated in a classical way using the available component reliability database and well known methods. On the other hand, different methods are required to evaluate the reliability of processes like thermohydraulics due to lack of adequate failure data. The research is ongoing worldwide on the reliability assessment of the passive systems and their integration into PSA, however consensus is not reached. Two of the most widely used methods are Reliability Evaluation of Passive Systems (REPAS) and Assessment of Passive System Reliability (APSRA). Both these methods characterize the uncertainties involved in the design and process parameters governing the function of the passive system. However, these methods differ in the quantification of passive system reliability. Inter comparison among different available methods provides useful insights into the strength and weakness of different methods. This paper highlights the results of the thermal hydraulic analysis of a typical passive isolation condenser system carried out using RELAP mode 3.2 computer code applying REPAS and APSRA methodologies. The failure surface is established for the passive system under consideration and system reliability has also been evaluated using these methods. Challenges involved in passive system reliabilities are identified, which require further attention in order to overcome the shortcomings of these

  18. 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

  19. 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.

  20. Electrostatic stabilizer for a passive magnetic bearing system

    Science.gov (United States)

    Post, Richard F.

    2015-11-24

    Electrostatic stabilizers are provided for passive bearing systems composed of annular magnets having a net positive stiffness against radial displacements and that have a negative stiffness for vertical displacements, resulting in a vertical instability. Further embodiments are shown of a radial electrostatic stabilizer geometry (using circuitry similar to that employed in the vertical stabilizer). This version is suitable for stabilizing radial (lateral) displacements of a rotor that is levitated by annular permanent magnets that are stable against vertical displacements but are unstable against radial displacements.

  1. Solar passive ceiling system. Final report. [Passive solar heating system with venetian blind reflectors and latent heat storage in ceiling

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, A.R.

    1980-01-01

    The construction of a 1200 square foot building, with full basement, built to be used as a branch library in a rural area is described. The primary heating source is a passive solar system consisting of a south facing window system. The system consists of: a set of windows located in the south facing wall only, composed of double glazed units; a set of reflectors mounted in each window which reflects sunlight up to the ceiling (the reflectors are similar to venetian blinds); a storage area in the ceiling which absorbs the heat from the reflected sunlight and stores it in foil salt pouches laid in the ceiling; and an automated curtain which automatically covers and uncovers the south facing window system. The system is totally passive and uses no blowers, pumps or other active types of heat distribution equipment. The building contains a basement which is normally not heated, and the north facing wall is bermed four feet high around the north side.

  2. A new passive radon-thoron discriminative measurement system

    International Nuclear Information System (INIS)

    Sciocchetti, G.; Sciocchetti, A.; Giovannoli, P.; DeFelice, P.; Cardellini, F.; Cotellessa, G.; Pagliari, M.

    2010-01-01

    A new passive radon-thoron discriminative measurement system has been developed for monitoring radon and thoron individually. It consists of a 'couple' of passive integrating devices with a CR39 nuclear track detector (NTD). The experimental prototype is based on the application of a new concept of NTD instrument developed at ENEA, named Alpha-PREM, acronym of piston radon exposure meter, which allows controlling the detector exposure with a patented sampling technique (Int. Eu. Pat. and US Pat.). The 'twin diffusion chambers system' was based on two A-PREM devices consisting of the standard device, named NTD-Rn, and a modified version, named NTD-Rn/Tn, which was set up to improve thoron sampling efficiency of the diffusion chamber, without changing the geometry and the start/stop function of the NTD-Rn device. Coupling devices fitted on each device allowed getting a system, which works as a double-chamber structure when deployed at the monitoring position. In this paper both technical and physical aspects are considered. (authors)

  3. A new passive radon-thoron discriminative measurement system.

    Science.gov (United States)

    Sciocchetti, G; Sciocchetti, A; Giovannoli, P; DeFelice, P; Cardellini, F; Cotellessa, G; Pagliari, M

    2010-10-01

    A new passive radon-thoron discriminative measurement system has been developed for monitoring radon and thoron individually. It consists of a 'couple' of passive integrating devices with a CR39 nuclear track detector (NTD). The experimental prototype is based on the application of a new concept of NTD instrument developed at ENEA, named Alpha-PREM, acronym of piston radon exposure meter, which allows controlling the detector exposure with a patented sampling technique (Int. Eu. Pat. and US Pat.). The 'twin diffusion chambers system' was based on two A-PREM devices consisting of the standard device, named NTD-Rn, and a modified version, named NTD-Rn/Tn, which was set up to improve thoron sampling efficiency of the diffusion chamber, without changing the geometry and the start/stop function of the NTD-Rn device. Coupling devices fitted on each device allowed getting a system, which works as a double-chamber structure when deployed at the monitoring position. In this paper both technical and physical aspects are considered.

  4. Performance behavior of the passive containment cooling system of a natural circulation BWR during postulated accident condition

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Nayak, A.K.; Jain, Vikas; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Passive systems are playing prominent role in the development of innovative nuclear reactor systems due to their simplicity, enhanced safety, reliability and economy. These systems are being considered for normal operation as well as accidental conditions of reactor following a postulated accident scenario to preclude the scenarios arising out of failure of active systems as well as to minimize the operator intervention. Indian innovative reactor AHWR being designed for thorium utilization employs various passive safety concepts. As containment is the ultimate barrier to the release of radioactivity, passive concepts are being employed in BWRs for minimize peak containment pressure in the containment during a postulated accident condition like LOCA. The concept of passive containment cooling system (PCCS) in the AHWR comprises of inclined tube heat exchangers located underneath an elevated pool that removes the heat from the steam-air atmosphere of containment following a LOCA by natural circulation of water inside the tubes. The steam condenses on the external surface of tubes of PCCS in addition to the wall of the containment which in turn depressurizes the containment. This paper deals with the performance assessment of PCCS of AHWR during a postulated design basis LOCA by using the best estimate code RELAP5/Mod3.2. (author)

  5. CERN safety system monitoring - SSM

    International Nuclear Information System (INIS)

    Hakulinen, T.; Ninin, P.; Valentini, F.; Gonzalez, J.; Salatko-Petryszcze, C.

    2012-01-01

    CERN SSM (Safety System Monitoring) is a system for monitoring state-of-health of the various access and safety systems of the CERN site and accelerator infrastructure. The emphasis of SSM is on the needs of maintenance and system operation with the aim of providing an independent and reliable verification path of the basic operational parameters of each system. Included are all network-connected devices, such as PLCs (local purpose control unit), servers, panel displays, operator posts, etc. The basic monitoring engine of SSM is a freely available system-monitoring framework Zabbix, on top of which a simplified traffic-light-type web-interface has been built. The web-interface of SSM is designed to be ultra-light to facilitate access from hand-held devices over slow connections. The underlying Zabbix system offers history and notification mechanisms typical of advanced monitoring systems. (authors)

  6. 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)

  7. Evaluation of Passive Containment Cooling System design of SMART built in GBS for ocean environment under the Fukushima Accident Condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Gil; Kim, Seong Gu; Lee, Jeong Ik; Lee, Kang Heon; Lee, Phil Seung [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The ONPP can be towed to the installation site after the SMART is constructed within the GBS in a dry dock. And, by incorporating IPSS (Integrated Passive Safety System), proposed by KAIST, in the ocean SMART the safety of the whole system will be significantly increased which can potentially eliminate any possibility of repeating Fukushima accident again. In recent years, KAIST research team is developing a very advanced concept of ocean NPPs which can avoid natural disasters while potentially increasing economy and enhancing public acceptance. Authors chose Korean small reactor, SMART as a reference system to demonstrate the feasibility of the proposed ocean NPP. Ocean SMART is mounted on GBS (Gravity Based Structure)

  8. Treatment of Passive Component Reliability in Risk-Informed Safety Margin Characterization FY 2010 Report

    Energy Technology Data Exchange (ETDEWEB)

    Robert W Youngblood

    2010-09-01

    The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, is founded on probabilistic characterizations of SSC performance.

  9. Design of Passive Decay Heat Removal System using Mercury Thermosyphon for SFR

    Energy Technology Data Exchange (ETDEWEB)

    You, Byung Hyun; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    In this study, thermosyphon application is suggested to accomplish the fully passive safety grade system and compactness of components via enhance the heat removal performance. A two-phase evaporating thermosyphon operates when the evaporator is heated, the working fluid start boiling, the vapor that is formed moves to the condenser, where it is condensed on the walls, giving up the heat of phase change to the cooling fluid. Gravity forces cause the condensate to condensed liquid flow to the evaporator again. These processes occur continuously, which causes transfer of heat from evaporator to condenser vice versa. After the thermal design and performance evaluation, the results were compared with the performance of conventional DRACS system. For the same amount of decay heat removal performance of PDRC system of KALIMER-600 mercury thermosyphon system can archive around 30∼50% of compactness. For the detailed design, improved analytical model and experimental data for the validation will be required to specify the new DHR system.

  10. Thermal hydraulic studies for passive heat transport systems relevant to advanced reactors

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Sharma, M.; Borgohain, A.; Srivastava, A.K.; Pilkhwal, D.S.; Maheshwari, N.K.

    2014-01-01

    Nuclear is the only non-green house gas generating power source that can replace fossil fuels and can be commercially deployed in large scale. However, the enormous developmental efforts and safety upgrades during the past six decades have somewhat eroded the economic competitiveness of water-cooled reactors which form the mainstay of the current nuclear power programme. Further, the introduction of the supercritical Rankine cycle and the gas turbine based advanced fuel cycles have enhanced the efficiency of fossil fired power plants (FPP) thereby reducing its greenhouse gas emissions. The ongoing development of ultra-supercritical and advanced ultra-supercritical turbines aims to further reduce the greenhouse gas emissions and economic competitiveness of FPPs. In the backdrop of these developments, the nuclear industry also initiated development of advanced nuclear power plants (NPP) with improved efficiency, sustainability and enhanced safety as the main goals. A review of the advanced reactor concepts being investigated currently reveals that excepting the SCWR, all other concepts use coolants other than water. The coolants used are lead, lead bismuth eutectic, liquid sodium, molten salts, helium and supercritical water. Besides, some of these are employing passive systems to transport heat from the core under normal operating conditions. In view of this, a study is in progress at BARC to examine the performance of simple passive systems using SC CO 2 , SCW, LBE and molten salts as the coolant. This paper deals with some of the recent results of these studies. The study focuses on the steady state, transient and stability behaviour of the passive systems with these coolants. (author)

  11. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

  12. Safety of mechanical devices. Safety of automation systems

    International Nuclear Information System (INIS)

    Pahl, G.; Schweizer, G.; Kapp, K.

    1985-01-01

    The paper deals with the classic procedures of safety engineering in the sectors mechanical engineering, electrical and energy engineering, construction and transport, medicine technology and process technology. Particular stress is laid on the safety of automation systems, control technology, protection of mechanical devices, reactor safety, mechanical constructions, transport systems, railway signalling devices, road traffic and protection at work in chemical plans. (DG) [de

  13. 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.

  14. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1979-01-01

    Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

  15. Optimal design of passive containment cooling system for innovative PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  16. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  17. Optimal design of passive containment cooling system for innovative PWR

    International Nuclear Information System (INIS)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon

    2017-01-01

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed

  18. Micro and nano devices in passive millimetre wave imaging systems

    Science.gov (United States)

    Appleby, R.

    2013-06-01

    The impact of micro and nano technology on millimetre wave imaging from the post war years to the present day is reviewed. In the 1950s whisker contacted diodes in mixers and vacuum tubes were used to realise both radiometers and radars but required considerable skill to realise the performance needed. Development of planar semiconductor devices such as Gunn and Schottky diodes revolutionised mixer performance and provided considerable improvement. The next major breakthrough was high frequency transistors based on gallium arsenide which were initially used at intermediate frequencies but later after further development at millimeter wave frequencies. More recently Monolithic Microwave Integrated circuits(MMICs) offer exceptional performance and the opportunity for innovative design in passive imaging systems. In the future the use of micro and nano technology will continue to drive system performance and we can expect to see integration of antennae, millimetre wave and sub millimetre wave circuits and signal processing.

  19. Some Passive Damping Sources on Flooring Systems besides the TMD

    DEFF Research Database (Denmark)

    Pedersen, Lars

    2010-01-01

    Impulsive loads and walking loads can generate problematic structural vibrations in flooring-systems. Measures that may be taken to mitigate the problem would often be to consider the implementation of a tuned mass damper or even more advanced vibration control technologies; this in order to add...... damping to the structure. Basically also passive humans on a floor act as a damping source, but it also turns out from doing system identification tests with a floor strip that a quite simple set-up installed on the floor (cheap and readily at hand) might do a good job in terms of reducing vertical floor...... vibrations for some floors. The paper describes the tests with the floor strip, and the results, in terms of dynamic floor behaviour, are compared with what would be expected had the floor instead been equipped with a tuned mass damper....

  20. Progress in passive solar energy systems. Volume 8. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, J.; Andrejko, D.A.

    1983-01-01

    This book presents the papers given at a conference sponsored by the US DOE, the Solar Energy Research Institute, SolarVision, Inc., and the Southern California Solar Energy Society. The topics considered at the conference included sizing solar energy systems for agricultural applications, a farm scale ethanol production plant, the EEC wind energy RandD program, the passive solar performance assessment of an earth-sheltered house, the ARCO 1 MW photovoltaic power plant, the performance of a dendritic web photovoltaic module, second generation point focused concentrators, linear fresnel lens concentrating photovoltaic collectors, photovoltaic conversion efficiency, amorphous silicon thin film solar cells, a photovoltaic system for a shopping center, photovoltaic power generation for the utility industry, spectral solar radiation, and the analysis of insolation data.

  1. Passive ventilation systems with heat recovery and night cooling

    DEFF Research Database (Denmark)

    Hviid, Christian Anker; Svendsen, Svend

    2008-01-01

    with little energy consumption and with satisfying indoor climate. The concept is based on using passive measures like stack and wind driven ventilation, effective night cooling and low pressure loss heat recovery using two fluid coupled water-to-air heat exchangers developed at the Technical University......In building design the requirements for energy consumption for ventilation, heating and cooling and the requirements for increasingly better indoor climate are two opposing factors. This paper presents the schematic layout and simulation results of an innovative multifunc-tional ventilation concept...... of Denmark. Through building integration in high performance offices the system is optimized to incorporate multiple functions like heating, cooling and ventilation, thus saving the expenses of separate cooling and heating systems. The simulation results are derived using the state-of-the-art building...

  2. Optimum radars and filters for the passive sphere system

    Science.gov (United States)

    Luers, J. K.; Soltes, A.

    1971-01-01

    Studies have been conducted to determine the influence of the tracking radar and data reduction technique on the accuracy of the meteorological measurements made in the 30 to 100 kilometer altitude region by the ROBIN passive falling sphere. A survey of accuracy requirements was made of agencies interested in data from this region of the atmosphere. In light of these requirements, various types of radars were evaluated to determine the tracking system most applicable to the ROBIN, and methods were developed to compute the errors in wind and density that arise from noise errors in the radar supplied data. The effects of launch conditions on the measurements were also examined. Conclusions and recommendations have been made concerning the optimum tracking and data reduction techniques for the ROBIN falling sphere system.

  3. The LHC personnel safety system

    International Nuclear Information System (INIS)

    Ninin, P.; Valentini, F.; Ladzinski, T.

    2011-01-01

    Large particle physics installations such as the CERN Large Hadron Collider require specific Personnel Safety Systems (PSS) to protect the personnel against the radiological and industrial hazards. In order to fulfill the French regulation in matter of nuclear installations, the principles of IEC 61508 and IEC 61513 standard are used as a methodology framework to evaluate the criticality of the installation, to design and to implement the PSS.The LHC PSS deals with the implementation of all physical barriers, access controls and interlock devices around the 27 km of underground tunnel, service zones and experimental caverns of the LHC. The system shall guarantee the absence of personnel in the LHC controlled areas during the machine operations and, on the other hand, ensure the automatic accelerator shutdown in case of any safety condition violation, such as an intrusion during beam circulation. The LHC PSS has been conceived as two separate and independent systems: the LHC Access Control System (LACS) and the LHC Access Safety System (LASS). The LACS, using off the shelf technologies, realizes all physical barriers and regulates all accesses to the underground areas by identifying users and checking their authorizations.The LASS has been designed according to the principles of the IEC 61508 and 61513 standards, starting from a risk analysis conducted on the LHC facility equipped with a standard access control system. It consists in a set of safety functions realized by a dedicated fail-safe and redundant hardware guaranteed to be of SIL3 class. The integration of various technologies combining electronics, sensors, video and operational procedures adopted to establish an efficient personnel safety system for the CERN LHC accelerator is presented in this paper. (authors)

  4. 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)

  5. 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)

  6. Safety in nuclear power systems

    International Nuclear Information System (INIS)

    Myers, L.C.

    1987-05-01

    This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents that have occurred to date. Details are also provided of Ontario Hydro's problems with Unit 2 at Pickering

  7. Firefighter Safety for PV Systems

    DEFF Research Database (Denmark)

    Mathe, Laszlo; Sera, Dezso; Spataru, Sergiu

    2015-01-01

    An important and highly discussed safety issue for photovoltaic (PV) systems is that as long as the PV panels are illuminated, a high voltage is present at the PV string terminals and cables between the string and inverters that is independent of the state of the inverter's dc disconnection switch...

  8. NASA aviation safety reporting system

    Science.gov (United States)

    1981-01-01

    Aviation safety reports that relate to loss of control in flight, problems that occur as a result of similar sounding alphanumerics, and pilot incapacitation are presented. Problems related to the go around maneuver in air carrier operations, and bulletins (and FAA responses to them) that pertain to air traffic control systems and procedures are included.

  9. Safety design guide for safety related systems for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.

  10. Safety design guide for safety related systems for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new

  11. Passive and active vibration isolation systems using inerter

    Science.gov (United States)

    Alujević, N.; Čakmak, D.; Wolf, H.; Jokić, M.

    2018-03-01

    This paper presents a theoretical study on passive and active vibration isolation schemes using inerter elements in a two degree of freedom (DOF) mechanical system. The aim of the work is to discuss basic capabilities and limitations of the vibration control systems at hand using simple and physically transparent models. Broad frequency band dynamic excitation of the source DOF is assumed. The purpose of the isolator system is to prevent vibration transmission to the receiving DOF. The frequency averaged kinetic energy of the receiving mass is used as the metric for vibration isolation quality. It is shown that the use of inerter element in the passive vibration isolation scheme can enhance the isolation effect. In the active case, a feedback disturbance rejection scheme is considered. Here, the error signal is the receiving body absolute velocity which is directly fed to a reactive force actuator between the source and the receiving bodies. In such a scheme, the so-called subcritical vibration isolation problems exist. These problems are characterised by the uncoupled natural frequency of the receiving body larger than the uncoupled natural frequency of the source body. In subcritical vibration isolation problems, the performance of the active control is limited by poor stability margins. This is because the stable feedback gain is restricted in a narrow range between a minimum and a maximum. However, with the inclusion of an inerter in the isolator, one of the two stability margins can be opened. This enables large, theoretically unlimited negative feedback gains and large active damping of the receiving body vibration. A simple expression for the required inertance is derived.

  12. The Daresbury personnel safety system

    International Nuclear Information System (INIS)

    Poole, D.E.; Ring, T.

    1989-01-01

    The personnel safety system designed for the SRS at Daresbury is a unified system covering the three accelerators of the source itself, the beamlines and the experimental stations. The system has also been applied to the experimental areas of the Nuclear Structure Facility, and is therefore established as a site standard. A dual guardline interlock module forms a building block for a relay based interlock system completely independent of the machine control system, although comprehensive monitoring of the system status via the control system computer is a feature. An outline of the design criteria adopted for the system is presented together with a more detailed description of the philosophy of the guardline logic and the way this is implemented in a standard modular form. The emphasis is on the design features of a modern microprocessor based variant of the original SRS system. Experience with the original system during build-up and operation of the SRS facility is described. 2 refs., 4 figs

  13. The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

    International Nuclear Information System (INIS)

    Hill, D.J.; Ragland, W.A.; Roglans, J.

    1998-01-01

    This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

  14. The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D J; Ragland, W A; Roglans, J

    1998-11-01

    This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10{sup -6} yr{sup -1}, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10{sup -6} yr{sup -1} and the contribution of seismic events is 1.7 10{sup -5} yr{sup -1}. Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

  15. The EBR-II probabilistic risk assessment lessons learned regarding passive safety

    International Nuclear Information System (INIS)

    Hill, D.J.; Ragland, W.A.; Roglans, J.

    1994-01-01

    This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

  16. Safety implications of control systems

    International Nuclear Information System (INIS)

    Smith, O.L.

    1983-01-01

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

  17. System safety education focused on industrial engineering

    Science.gov (United States)

    Johnston, W. L.; Morris, R. S.

    1971-01-01

    An educational program, designed to train students with the specific skills needed to become safety specialists, is described. The discussion concentrates on application, selection, and utilization of various system safety analytical approaches. Emphasis is also placed on the management of a system safety program, its relationship with other disciplines, and new developments and applications of system safety techniques.

  18. Availability analysis of the AP600 passive core cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, M [National Atomic Energy Research Agency, Yogyakarta (Indonesia); Subki, I R [BATAN Head Office, Jakarta (Indonesia); Canton, M H [Westinghouse Electric Corp. (United States)

    1996-12-01

    The reliability analysis of the AP600 Passive Core Cooling System (PXS) has been done. The fault tree analysis method was used for the quantitative analysis. The PXS can be grouped to several sub-systems i.e.: Reactor Coolant System (RCS) Injection Subsystem, Emergency Core Decay Heat Removal Subsystem, and Containment Sump pH Control Subsystem. The quantitative analysis results indicates that the system unavailability is highly dependent on the valves configuration of the Automatic Depressurization System (ADS). If the ADS valves is arranged in Option-1, the system unavailability is 2.347E-03, this means that the yearly contribution to plant down time can be estimated to be about 20.56 hours per year. Whereas, by using Option-2 of fourth stage ADS valves, the system unavailability is reduced to be 9.877E-04 or 8.65 hours per year and this value is consistent with the allocated goal value (8.0 hours per year). The ADS contributes 66.89% to the system unavailability if it is arranged in Option-1, and will reduced to be about 21.21% if its fourth stages are arranged in Option-2. If the ADS is not included as a subsystem of the PXS (relocate to RCS as a subsystem of RCS), then the PXS unavailability will be reduced to about 7.784E-04 or 6.82 hours per year; this is less then the allocated goal value. The major contributors to the system unavailability are mostly dominated by Stage-4 ADS valves (air piston operated valves and squib valves), inservice testing valves of ADS (solenoid operated valves), solenoid valves of Nitrogen Supply to Accumulator, and Passive Residual Heat Removal actuation valves (air operated valves). It is recommended that those valves be analyzed more detail to gain the improvement in its reliability. It is also recommended that the fourth stage of ADS valves should be arranged according to Option-2, i.e. one 10-inch normally open motor operated gate valve in series with one 10-inch normally closed squib valve. (author). 13 refs, 3 figs, 3 tabs.

  19. How could intelligent safety transport systems enhance safety ?

    NARCIS (Netherlands)

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

    2017-01-01

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

  20. Renewable building energy systems and passive human comfort solutions

    Energy Technology Data Exchange (ETDEWEB)

    Omer, Abdeen Mustafa [17 Juniper Court, Forest Road West, Nottingham NG7 4EU (United Kingdom)

    2008-08-15

    With environmental protection posing as the number one global problem, man has no choice but to reduce his energy consumption. One way to accomplish this is to resort to passive and low-energy systems to maintain thermal comfort in buildings. The conventional and modern designs of wind towers can successfully be used in hot arid regions to maintain thermal comfort (with or without the use of ceiling fans) during all hours of the cooling season, or a fraction of it. Climatic design is one of the best approaches to reduce the energy cost in buildings. Proper design is the first step of defence against the stress of the climate. Buildings should be designed according to the climate of the site, reducing the need for mechanical heating or cooling. Hence maximum natural energy can be used for creating a pleasant environment inside the built envelope. Technology and industry progress in the last decade diffused electronic and informatics' devices in many human activities, and also in building construction. The utilisation and operating opportunities components, increase the reduction of heat losses by varying the thermal insulation, optimise the lighting distribution with louver screens and operate mechanical ventilation for coolness in indoor spaces. In addition to these parameters the intelligent envelope can act for security control and became an important part of the building domotic revolution. Application of simple passive cooling measure is effective in reducing the cooling load of buildings in hot and humid climates. Fourty-three percent reductions can be achieved using a combination of well-established technologies such as glazing, shading, insulation, and natural ventilation. More advanced passive cooling techniques such as roof pond, dynamic insulation, and evaporative water jacket need to be considered more closely. The building sector is a major consumer of both energy and materials worldwide, and that consumption is increasing. Most industrialised

  1. Nuclear power systems: Their safety

    International Nuclear Information System (INIS)

    Myers, L.C.

    1993-01-01

    Mankind utilizes energy in many forms and from a variety of sources. Canada is one of a growing number of countries which have chosen to embrace nuclear-electric generation as a component of their energy systems. As of August 1992 there were 433 power reactors operating in 35 countries and accounting for more than 15% of the world's production of electricity. In 1992, thirteen countries derived at least 25% of their electricity from nuclear units, with France leading at nearly 70%. In the same year, Canada produced about 16% of its electricity from nuclear units. Some 68 power reactors are under construction in 16 countries, enough to expand present generating capacity by close to 20%. No human endeavour carries the guarantee of perfect safety and the question of whether or not nuclear-electric generation represents an 'acceptable' risk to society has long been vigorously debated. Until the events of late April 1986, nuclear safety had indeed been an issue for discussion, for some concern, but not for alarm. The accident at the Chernobyl reactor in the USSR has irrevocably changed all that. This disaster brought the matter of nuclear safety back into the public mind in a dramatic fashion. This paper discusses the issue of safety in complex energy systems and provides brief accounts of some of the most serious reactor accidents which have occurred to date. (author). 7 refs

  2. Radiation safety systems at the NSLS

    International Nuclear Information System (INIS)

    Dickinson, T.

    1987-04-01

    This report describes design principles that were used to establish the radiation safety systems at the National Synchrotron Light Source. The author described existing safety systems and the history of partial system failures. 1 fig

  3. How to effectively compute the reliability of a thermal-hydraulic nuclear passive system

    International Nuclear Information System (INIS)

    Zio, E.; Pedroni, N.

    2011-01-01

    Research highlights: → Optimized LS is the preferred choice for failure probability estimation. → Two alternative options are suggested for uncertainty and sensitivity analyses. → SS for simulation codes requiring seconds or minutes to run. → Regression models (e.g., ANNs) for simulation codes requiring hours or days to run. - Abstract: The computation of the reliability of a thermal-hydraulic (T-H) passive system of a nuclear power plant can be obtained by (i) Monte Carlo (MC) sampling the uncertainties of the system model and parameters, (ii) computing, for each sample, the system response by a mechanistic T-H code and (iii) comparing the system response with pre-established safety thresholds, which define the success or failure of the safety function. The computational effort involved can be prohibitive because of the large number of (typically long) T-H code simulations that must be performed (one for each sample) for the statistical estimation of the probability of success or failure. The objective of this work is to provide operative guidelines to effectively handle the computation of the reliability of a nuclear passive system. Two directions of computation efficiency are considered: from one side, efficient Monte Carlo Simulation (MCS) techniques are indicated as a means to performing robust estimations with a limited number of samples: in particular, the Subset Simulation (SS) and Line Sampling (LS) methods are identified as most valuable; from the other side, fast-running, surrogate regression models (also called response surfaces or meta-models) are indicated as a valid replacement of the long-running T-H model codes: in particular, the use of bootstrapped Artificial Neural Networks (ANNs) is shown to have interesting potentials, including for uncertainty propagation. The recommendations drawn are supported by the results obtained in an illustrative application of literature.

  4. IAEA Safety Standards on Management Systems and Safety Culture

    International Nuclear Information System (INIS)

    Persson, Kerstin Dahlgren

    2007-01-01

    The IAEA has developed a new set of Safety Standard for applying an integrated Management System for facilities and activities. The objective of the new Safety Standards is to define requirements and provide guidance for establishing, implementing, assessing and continually improving a Management System that integrates safety, health, environmental, security, quality and economic related elements to ensure that safety is properly taken into account in all the activities of an organization. With an integrated approach to management system it is also necessary to include the aspect of culture, where the organizational culture and safety culture is seen as crucial elements of the successful implementation of this management system and the attainment of all the goals and particularly the safety goals of the organization. The IAEA has developed a set of service aimed at assisting it's Member States in establishing. Implementing, assessing and continually improving an integrated management system. (author)

  5. Study of passive residual heat removal system of a modular small PWR reactor

    International Nuclear Information System (INIS)

    Araujo, Nathália N.; Su, Jian

    2017-01-01

    This paper presents a study on the passive residual heat removal system (PRHRS) of a small modular nuclear reactor (SMR) of 75MW. More advanced nuclear reactors, such as generation III + and IV, have passive safety systems that automatically go into action in order to prevent accidents. The purpose of the PRHRS is to transfer the decay heat from the reactor's nuclear fuel, keeping the core cooled after the plant has shut down. It starts operating in the event of fall of power supply to the nuclear station, or in the event of an unavailability of the steam generator water supply system. Removal of decay heat from the core of the reactor is accomplished by the flow of the primary refrigerant by natural circulation through heat exchangers located in a pool filled with water located above the core. The natural circulation is caused by the density gradient between the reactor core and the pool. A thermal and comparative analysis of the PRHRS was performed consisting of the resolution of the mass conservation equations, amount of movement and energy and using incompressible fluid approximations with the Boussinesq approximation. Calculations were performed with the aid of Mathematica software. A design of the heat exchanger and the cooling water tank was done so that the core of the reactor remained cooled for 72 hours using only the PRHRS

  6. Integrated therapy safety management system.

    Science.gov (United States)

    Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

    2013-09-01

    The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an 'integrated therapy safety management' is drafted. This concept could serve as a basis to improve resilience. The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for 'integrated therapy safety management'. The concept is applied by way of example for the 'medication process' to demonstrate its practical implementation. The 'integrated therapy safety management' is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of 'bridge managers'. 'Bridge managers' anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the 'bridge managers' and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. © 2013 The Authors. British Journal of Clinical Pharmacology © 2013 The British Pharmacological Society.

  7. Integrated therapy safety management system

    Science.gov (United States)

    Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

    2013-01-01

    Aims The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an ‘integrated therapy safety management’ is drafted. This concept could serve as a basis to improve resilience. Methods The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for ‘integrated therapy safety management’. The concept is applied by way of example for the ‘medication process’ to demonstrate its practical implementation. Results The ‘integrated therapy safety management’ is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of ‘bridge managers’. ‘Bridge managers’ anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the ‘bridge managers’ and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. Conclusions The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. PMID:24007448

  8. The ATLAS Detector Safety System

    CERN Multimedia

    Helfried Burckhart; Kathy Pommes; Heidi Sandaker

    The ATLAS Detector Safety System (DSS) has the mandate to put the detector in a safe state in case an abnormal situation arises which could be potentially dangerous for the detector. It covers the CERN alarm severity levels 1 and 2, which address serious risks for the equipment. The highest level 3, which also includes danger for persons, is the responsibility of the CERN-wide system CSAM, which always triggers an intervention by the CERN fire brigade. DSS works independently from and hence complements the Detector Control System, which is the tool to operate the experiment. The DSS is organized in a Front- End (FE), which fulfills autonomously the safety functions and a Back-End (BE) for interaction and configuration. The overall layout is shown in the picture below. ATLAS DSS configuration The FE implementation is based on a redundant Programmable Logical Crate (PLC) system which is used also in industry for such safety applications. Each of the two PLCs alone, one located underground and one at the s...

  9. Synchronous Design and Test of Distributed Passive Radar Systems Based on Digital Broadcasting and Television

    Directory of Open Access Journals (Sweden)

    Wan Xianrong

    2017-02-01

    Full Text Available Digital broadcasting and television are important classes of illuminators of opportunity for passive radars. Distributed and multistatic structure are the development trends for passive radars. Most modern digital broadcasting and television systems work on a network, which not only provides a natural condition to distributed passive radar but also puts forward higher requirements on the design of passive radar systems. Among those requirements, precise synchronization among the receivers and transmitters as well as among multiple receiving stations, which mainly involves frequency and time synchronization, is the first to be solved. To satisfy the synchronization requirements of distributed passive radars, a synchronization scheme based on GPS is presented in this paper. Moreover, an effective scheme based on the China Mobile Multimedia Broadcasting signal is proposed to test the system synchronization performance. Finally, the reliability of the synchronization design is verified via the distributed multistatic passive radar experiments.

  10. Modeling illumination performance of plastic optical fiber passive daylighting system

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, F; Ahmad, A [Universiti Teknologi MARA, Shah Alam (Malaysia). Faculty of Electrical Engineering; Ahmed, A Z [Universiti Teknologi MARA, Shah Alam (Malaysia). Bureau of Reseaarch and Consultancy

    2006-12-15

    of the most direct methods of utilizing solar energy for energy conservation is to bring natural light indoors to light up an area. This paper reports on the investigation of the feasibility to utilize large core optical fibers to convey and distribute solar light passively throughout residential or commercial structures. The focus of this study is on the mathematical modeling of the illumination performance and the light transmission efficiency of solid core end light fiber for optical day lighting systems. The Meatball simulations features the optical fiber transmittance for glass and plastic fibers, illumination performance over lengths of plastic end-lit fiber, spectral transmission, light intensity loss through the large diameter solid core optical fibers as well as the transmission efficiency of the optical fiber itself. It was found that plastic optical fiber has less transmission loss over the distance of the fiber run which clearly shows that the Plastic Optical Fiber should be optimized for emitting visible light. The findings from the analysis on the performance of large diameter optical fibers for day lighting systems seems feasible for energy efficient lighting system in commercial or residential buildings.

  11. Modeling illumination performance of plastic optical fiber passive daylighting system

    International Nuclear Information System (INIS)

    Sulaiman, F.; Ahmad, A.; Ahmed, A.Z.

    2006-01-01

    One of the most direct methods of utilizing solar energy for energy conservation is to bring natural light indoors to light up an area. This paper reports on the investigation of the feasibility to utilize large core optical fibers to convey and distribute solar light passively throughout residential or commercial structures. The focus of this study is on the mathematical modeling of the illumination performance and the light transmission efficiency of solid core end light fiber for optical day lighting systems. The Meatball simulations features the optical fiber transmittance for glass and plastic fibers, illumination performance over lengths of plastic end-lit fiber, spectral transmission, light intensity loss through the large diameter solid core optical fibers as well as the transmission efficiency of the optical fiber itself. It was found that plastic optical fiber has less transmission loss over the distance of the fiber run which clearly shows that the Plastic Optical Fiber should be optimized for emitting visible light. The findings from the analysis on the performance of large diameter optical fibers for day lighting systems seems feasible for energy efficient lighting system in commercial or residential buildings

  12. Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

    1991-01-01

    A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

  13. Patient-Centered Robot-Aided Passive Neurorehabilitation Exercise Based on Safety-Motion Decision-Making Mechanism

    Directory of Open Access Journals (Sweden)

    Lizheng Pan

    2017-01-01

    Full Text Available Safety is one of the crucial issues for robot-aided neurorehabilitation exercise. When it comes to the passive rehabilitation training for stroke patients, the existing control strategies are usually just based on position control to carry out the training, and the patient is out of the controller. However, to some extent, the patient should be taken as a “cooperator” of the training activity, and the movement speed and range of the training movement should be dynamically regulated according to the internal or external state of the subject, just as what the therapist does in clinical therapy. This research presents a novel motion control strategy for patient-centered robot-aided passive neurorehabilitation exercise from the point of the safety. The safety-motion decision-making mechanism is developed to online observe and assess the physical state of training impaired-limb and motion performances and regulate the training parameters (motion speed and training rage, ensuring the safety of the supplied rehabilitation exercise. Meanwhile, position-based impedance control is employed to realize the trajectory tracking motion with interactive compliance. Functional experiments and clinical experiments are investigated with a healthy adult and four recruited stroke patients, respectively. The two types of experimental results demonstrate that the suggested control strategy not only serves with safety-motion training but also presents rehabilitation efficacy.

  14. Development of a hybrid safety system: Actuation of the secondary automatic depressurization system at an early stage

    International Nuclear Information System (INIS)

    Nishimoto, Masae; Umezawa, Shigemitsu; Okabe, Kazuharu; Matsuoka, Tsuyoshi

    1996-01-01

    A Hybrid Safety System, which is an optimum combination of active and passive safety systems, has been developed in order to improve the safety, reliability and economic features of the next generation of PWRs. The passive safety systems include Automatic primary Depressurization System (ADS), Secondary Automatic Depressurization System (SADS), advanced accumulators, gravity injection system and so on. In this study the authors have improved the actuation logic of the passive safety systems. The original logic in the previous study actuates ADS at an early stage of an event such as a Loss-of-Coolant Accident (LOCA), and this is followed by the actuation of SADS. In this study they divide SADS into two systems. The first, small SADS, uses small valves corresponding to the relief valves of the conventional PWR plants. The second, large SADS, corresponds to the original SADS using multiple valves of large capacity. With the new logic, the passive systems are actuated during a typical small LOCA. Small LOCA analyses using several break areas were performed for a 1,400 MWe PWR plant with a Hybrid Safety System. The results predict that core uncovery does not occur in the case of a relatively small break area and that core heat removal during a small LOCA is improved in comparison with the analyses for conventional PWR plants, where the secondary pressure remains higher during the event. The results also predict that this new logic make it possible to reduce the ADS valve size and the actuation pressure setpoint of the passive safety systems

  15. A passive wireless ultrasound pitch–catch system

    International Nuclear Information System (INIS)

    Zahedi, F; Yao, J; Huang, H

    2015-01-01

    This paper exploits amplitude modulation and demodulation to achieve a passive wireless ultrasound pitch–catch system consisting of a wireless interrogator and a combination of a wireless actuator and a sensor mounted on a structure. The wireless interrogator operates in two modes, i.e. the generation and sensing modes. At the generation mode, the interrogator transmits two microwave signals; one is amplitude modulated with the ultrasound excitation signal while the other is a continuous-wave carrier signal. Once received by the wireless actuator, the amplitude modulated signal is demodulated using the carrier signal to recover the ultrasound excitation signal, which is then supplied to a piezoelectric wafer actuator for ultrasound generation. Subsequently, the interrogator is switched to the sensing mode by transmitting a carrier signal with a different frequency. Once received by the wireless sensor, this carrier signal is modulated with the ultrasound sensing signal acquired by the piezoelectric wafer sensor to produce an amplitude modulated microwave signal, which can then be wirelessly transmitted and demodulated by the interrogator to recover the original ultrasound sensing signal. The principle and implementation of the wireless ultrasound pitch–catch system as well as the data processing of the wirelessly received sensing signal are described. Experiment results validating wireless ultrasound generation and sensing from a distance of 0.5 m are presented. (paper)

  16. Foundations for Improvements to Passive Detection Systems - Final Report

    International Nuclear Information System (INIS)

    Labov, S E; Pleasance, L; Sokkappa, P; Craig, W; Chapline, G; Frank, M; Gronberg, J; Jernigan, J G; Johnson, S; Kammeraad, J; Lange, D; Meyer, A; Nelson, K; Pohl, B; Wright, D; Wurtz, R

    2004-01-01

    This project explores the scientific foundation and approach for improving passive detection systems for plutonium and highly enriched uranium in real applications. Sources of gamma-ray radiation of interest were chosen to represent a range of national security threats, naturally occurring radioactive materials, industrial and medical radiation sources, and natural background radiation. The gamma-ray flux emerging from these sources, which include unclassified criticality experiment configurations as surrogates for nuclear weapons, were modeled in detail. The performance of several types of gamma-ray imaging systems using Compton scattering were modeled and compared. A mechanism was created to model the combine sources and background emissions and have the simulated radiation ''scene'' impinge on a model of a detector. These modeling tools are now being used in various projects to optimize detector performance and model detector sensitivity in complex measuring environments. This study also developed several automated algorithms for isotope identification from gamma-ray spectra and compared these to each other and to algorithms already in use. Verification testing indicates that these alternative isotope identification algorithms produced less false positive and false negative results than the ''GADRAS'' algorithms currently in use. In addition to these algorithms that used binned spectra, a new approach to isotope identification using ''event mode'' analysis was developed. Finally, a technique using muons to detect nuclear material was explored

  17. HTS thin films: Passive microwave components and systems integration issues

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, F.A.; Chorey, C.M.; Bhasin, K.B. [National Aeronautics and Space Administration, Cleveland, OH (United States)

    1994-12-31

    The excellent microwave properties of the High-Temperature-Superconductors (HTS) have been amply demonstrated in the laboratory by techniques such as resonant cavity, power transmission and microstrip resonator measurements. The low loss and high Q passive structures made possible with HTS, present attractive options for applications in commercial, military and space-based systems. However, to readily insert HTS into these systems improvement is needed in such areas as repeatability in the deposition and processing of the HTS films, metal-contact formation, wire bonding, and overall film endurance to fabrication and assembly procedures. In this paper we present data compiled in our lab which illustrate many of the problems associated with these issues. Much of this data were obtained in the production of a space qualified hybrid receiver-downconverter module for the Naval Research Laboratory`s High Temperature Superconductivity Space Experiment II (HTSSE-II). Examples of variations observed in starting films and finished circuits will be presented. It is shown that under identical processing the properties of the HTS films can degrade to varying extents. Finally, we present data on ohmic contacts and factors affecting their adhesion to HTS films, strength of wire bonds made to such contacts, and aging effects.

  18. TRAC analysis of passive containment cooling system performance

    International Nuclear Information System (INIS)

    Arai, Kenji; Kataoka, Kazuyoshi; Nagasaka, Hideo

    1993-01-01

    A passive containment cooling system (PCCS) is a promising concept to improve the reliability of decay heat removal during an accident. Toshiba has carried out analytical studies for PCCS development in addition to experimental studies, using a best estimate thermal hydraulic computer code TRAC. In order to establish an analytical model for the PCCS performance analysis, it is necessary for the analytical model to be qualified against experimental results and thoroughly address the phenomena important for PCCS performance analysis. In this paper, the TRAC qualification for PCCS application is reported. A TRAC model has been verified against a drain line break simulation test conducted at the PCCS integral test facility, GIRAFFE. The result shows that the TRAC model can accurately predict the major system response and the PCCS performance in the drain line break test. In addition, the results of several sensitivity analyses, showing various points concerning the modeling in the PCCS performance analysis, have been reported. The analyses have been carried out for the SBWR and the analytical points are closely related to important phenomena which can affect PCCS performance

  19. HTS thin films: Passive microwave components and systems integration issues

    International Nuclear Information System (INIS)

    Miranda, F.A.; Chorey, C.M.; Bhasin, K.B.

    1994-01-01

    The excellent microwave properties of the High-Temperature-Superconductors (HTS) have been amply demonstrated in the laboratory by techniques such as resonant cavity, power transmission and microstrip resonator measurements. The low loss and high Q passive structures made possible with HTS, present attractive options for applications in commercial, military and space-based systems. However, to readily insert HTS into these systems improvement is needed in such areas as repeatability in the deposition and processing of the HTS films, metal-contact formation, wire bonding, and overall film endurance to fabrication and assembly procedures. In this paper we present data compiled in our lab which illustrate many of the problems associated with these issues. Much of this data were obtained in the production of a space qualified hybrid receiver-downconverter module for the Naval Research Laboratory's High Temperature Superconductivity Space Experiment II (HTSSE-II). Examples of variations observed in starting films and finished circuits will be presented. It is shown that under identical processing the properties of the HTS films can degrade to varying extents. Finally, we present data on ohmic contacts and factors affecting their adhesion to HTS films, strength of wire bonds made to such contacts, and aging effects

  20. On-Chip Correlator for Passive Wireless SAW Multisensor Systems

    Directory of Open Access Journals (Sweden)

    Liqiang Xie

    2016-01-01

    Full Text Available For decoding the asynchronous superposition of response signals from different sensors, it is a challenge to achieve correlation in a code division multiplexing (CDM based passive wireless surface acoustic wave (SAW multisensor system. Therefore, an on-chip correlator scheme is developed in this paper. In contrast to conventional CDM-based systems, this novel scheme enables the correlations to be operated at the SAW sensors, instead of the reader. Thus, the response signals arriving at the reader are the result of cross-correlation on the chips. It is then easy for the reader to distinguish the sensor that is matched with the interrogating signal. The operation principle, signal analysis, and simulation of the novel scheme are described in the paper. The simulation results show the response signals from the correlations of the sensors. A clear spike pulse is presented in the response signals, when a sensor code is matched with the interrogating code. Simulations verify the feasibility of the on-chip correlator concept.

  1. 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)

  2. Characterization of Passive Spectral Regrowth in Radio Frequency Systems

    Science.gov (United States)

    2013-01-01

    as using RF absorber and Faraday cages around sensi- tive spots. To ensure maximum radiated isolation, each cable or component should be shielded...nonlinear effects of spectral-regrowth-generating phenomena on an RF signal. Detection of low-level passive spectral regrowth close in frequency to a...experimentally and analytically characterize the nonlinear effects of spectral- regrowth-generating phenomena on an RF signal. Detection of low-level passive

  3. Safety performance monitoring of autonomous marine systems

    International Nuclear Information System (INIS)

    Thieme, Christoph A.; Utne, Ingrid B.

    2017-01-01

    The marine environment is vast, harsh, and challenging. Unanticipated faults and events might lead to loss of vessels, transported goods, collected scientific data, and business reputation. Hence, systems have to be in place that monitor the safety performance of operation and indicate if it drifts into an intolerable safety level. This article proposes a process for developing safety indicators for the operation of autonomous marine systems (AMS). The condition of safety barriers and resilience engineering form the basis for the development of safety indicators, synthesizing and further adjusting the dual assurance and the resilience based early warning indicator (REWI) approaches. The article locates the process for developing safety indicators in the system life cycle emphasizing a timely implementation of the safety indicators. The resulting safety indicators reflect safety in AMS operation and can assist in planning of operations, in daily operational decision-making, and identification of improvements. Operation of an autonomous underwater vehicle (AUV) exemplifies the process for developing safety indicators and their implementation. The case study shows that the proposed process leads to a comprehensive set of safety indicators. It is expected that application of the resulting safety indicators consequently will contribute to safer operation of current and future AMS. - Highlights: • Process for developing safety indicators for autonomous marine systems. • Safety indicators based on safety barriers and resilience thinking. • Location of the development process in the system lifecycle. • Case study on AUV demonstrating applicability of the process.

  4. 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)

  5. Passive magnetic shielding in MRI-Linac systems

    Science.gov (United States)

    Whelan, Brendan; Kolling, Stefan; Oborn, Brad M.; Keall, Paul

    2018-04-01

    Passive magnetic shielding refers to the use of ferromagnetic materials to redirect magnetic field lines away from vulnerable regions. An application of particular interest to the medical physics community is shielding in MRI systems, especially integrated MRI-linear accelerator (MRI-Linac) systems. In these systems, the goal is not only to minimize the magnetic field in some volume, but also to minimize the impact of the shield on the magnetic fields within the imaging volume of the MRI scanner. In this work, finite element modelling was used to assess the shielding of a side coupled 6 MV linac and resultant heterogeneity induced within the 30 cm diameter of spherical volume (DSV) of a novel 1 Tesla split bore MRI magnet. A number of different shield parameters were investigated; distance between shield and magnet, shield shape, shield thickness, shield length, openings in the shield, number of concentric layers, spacing between each layer, and shield material. Both the in-line and perpendicular MRI-Linac configurations were studied. By modifying the shield shape around the linac from the starting design of an open ended cylinder, the shielding effect was boosted by approximately 70% whilst the impact on the magnet was simultaneously reduced by approximately 10%. Openings in the shield for the RF port and beam exit were substantial sources of field leakage; however it was demonstrated that shielding could be added around these openings to compensate for this leakage. Layering multiple concentric shield shells was highly effective in the perpendicular configuration, but less so for the in-line configuration. Cautious use of high permeability materials such as Mu-metal can greatly increase the shielding performance in some scenarios. In the perpendicular configuration, magnetic shielding was more effective and the impact on the magnet lower compared with the in-line configuration.

  6. 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.)

  7. An optimized Line Sampling method for the estimation of the failure probability of nuclear passive systems

    International Nuclear Information System (INIS)

    Zio, E.; Pedroni, N.

    2010-01-01

    The quantitative reliability assessment of a thermal-hydraulic (T-H) passive safety system of a nuclear power plant can be obtained by (i) Monte Carlo (MC) sampling the uncertainties of the system model and parameters, (ii) computing, for each sample, the system response by a mechanistic T-H code and (iii) comparing the system response with pre-established safety thresholds, which define the success or failure of the safety function. The computational effort involved can be prohibitive because of the large number of (typically long) T-H code simulations that must be performed (one for each sample) for the statistical estimation of the probability of success or failure. In this work, Line Sampling (LS) is adopted for efficient MC sampling. In the LS method, an 'important direction' pointing towards the failure domain of interest is determined and a number of conditional one-dimensional problems are solved along such direction; this allows for a significant reduction of the variance of the failure probability estimator, with respect, for example, to standard random sampling. Two issues are still open with respect to LS: first, the method relies on the determination of the 'important direction', which requires additional runs of the T-H code; second, although the method has been shown to improve the computational efficiency by reducing the variance of the failure probability estimator, no evidence has been given yet that accurate and precise failure probability estimates can be obtained with a number of samples reduced to below a few hundreds, which may be required in case of long-running models. The work presented in this paper addresses the first issue by (i) quantitatively comparing the efficiency of the methods proposed in the literature to determine the LS important direction; (ii) employing artificial neural network (ANN) regression models as fast-running surrogates of the original, long-running T-H code to reduce the computational cost associated to the

  8. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  9. SULEU NTP Core with Passive Reactivity Control and Enhanced Submersion Safety

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Eades, Michael J [The Ohio State University, Ohio (United States)

    2016-05-15

    In this summary, SULEU has been adapted to implement some of the latest developments of LEUNTP design efforts. These include the implementation of a rapid depletion burnable absorber to flatten the reactivity profile during operation and the addition of a lower axial reflector to help minimize the reactivity increase during the full submersion criticality accident. The purpose of this study is to show the state of current LEU-NTP designs in terms of resolving key issues such as minimizing control drum usage and resolving the full submersion criticality accident. Future work will include integrating the rapid depletion poison with other passive reactivity control devices (such as hydrogen density in the tie-tubes) and developing additional systems for mitigating the full submersion criticality accident. It is widely acknowledged that nuclear thermal propulsion (NTP) is an enabling technology for manned missions to Mars and other locations beyond low-Earth orbit. Without nuclear thermal propulsion, manned space travel will be severely limited by the propellant requirements of chemical propulsion and significantly longer travel times of electric propulsion. While the performance superiority of NTP is clear, its implementation has been to date unsuccessful due to the significant costs of development, implementation, and regulations associated with the heritage NTP designs. These new systems take heritage designs and experimental results and adapt them to use LEU fuel with minimum impact on the heritage system. This is done in order to ensure their continued relevance with existing NTP research efforts and enable their rapid implementation into existing NASA efforts for human Mars mission planning. Of the current baseline NTP designs being studied, this paper concerns itself with the improvement of the Superb Use of Low Enriched (SULEU) core.

  10. Systems engineered health and safety criteria for safety analysis reports

    International Nuclear Information System (INIS)

    Beitel, G.A.; Morcos, N.

    1993-01-01

    The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

  11. 77 FR 70409 - System Safety Program

    Science.gov (United States)

    2012-11-26

    ...-0060, Notice No. 2] 2130-AC31 System Safety Program AGENCY: Federal Railroad Administration (FRA... rulemaking (NPRM) published on September 7, 2012, FRA proposed regulations to require commuter and intercity passenger railroads to develop and implement a system safety program (SSP) to improve the safety of their...

  12. Comprehensive Lifecycle for Assuring System Safety

    Science.gov (United States)

    Knight, John C.; Rowanhill, Jonathan C.

    2017-01-01

    CLASS is a novel approach to the enhancement of system safety in which the system safety case becomes the focus of safety engineering throughout the system lifecycle. CLASS also expands the role of the safety case across all phases of the system's lifetime, from concept formation to decommissioning. As CLASS has been developed, the concept has been generalized to a more comprehensive notion of assurance becoming the driving goal, where safety is an important special case. This report summarizes major aspects of CLASS and contains a bibliography of papers that provide additional details.

  13. Assessment of a potential rapid condensation induced water hammer in a passive auxiliary feedwater system

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Shin, Byung Soo; Do, Kyu Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Moody, Frederick J. [General Electric (Retired), CA (United States)

    2012-10-15

    A passive auxiliary feedwater system (PAFS) which is incorporated in the APR+ system is a kind of closed natural circulation loop. The PAFS has no operating functions during normal plant operation, but it has a dedicated safety function of the residual heat removal following initiating events, including the unlikely event of the most limiting single failure occurring coincident with a loss of offsite power, when the feedwater system becomes inoperable or unavailable. Even in the unlikely event of a station blackout, the isolation valves can be opened either by DC power or manual operation and then the PAFS can also provide adequate condensate to the steam generator (SG). The PAFS piping in the vicinity of each of the two SGs is designed to minimize the potential for destructive water hammer during start up operation by setting the stroke time for full close or full open of the condensate isolation valves upon receipt of a passive auxiliary feedwater actuation signal. The temperature of the stagnant condensate water and its surrounding tubes and piping during the reactor normal operation modes may fall to the ambient temperature. A possible concern is the introduction of saturated steam into the PAFS recirculation pipe downstream of the PCHX in the beginning of the PAFS operation. Although the steam introduction rate is expected to be slow, a rapid condensation rate is expected due to the initial cold surrounding temperature in the pipe, which could result in a localized pressure reduction and the propagation of decompression and velocity disturbances into the condensate water leg, which might cause the sudden closure of check valves and associated water hammer. Thus, it is requisite for the licensing review of the PAFS design to confirm if destructive water hammers will not be produced due to such rapid condensation induced decompressions in the system. This paper addresses an assessment of the potential local decompressions which could result from the steam

  14. An innovative pool with a passive heat removal system

    International Nuclear Information System (INIS)

    Vitale Di Maio, Damiano; Naviglio, Antonio; Giannetti, Fabio; Manni, Fabio

    2012-01-01

    Heat removal systems are of primary importance in several industrial processes. As heat sink, a water pool or atmospheric air may be selected. The first solution takes advantage of high heat transfer coefficient with water but it requires active systems to maintain a constant water level; the second solution takes benefit from the unlimited heat removal capacity by air, but it requires a larger heat exchanger to compensate the lower heat transfer coefficient. In NPPs (nuclear power plants) during a nuclear reactor shutdown, as well as in some chemical plants to control runaway reactions, it is possible to use an innovative heat sink that joins the advantages of the two previous solutions. This solution is based on a special heat exchanger submerged in a water pool designed so that when heat removal is requested, active systems are not required to maintain the water level; due to the special design, when the pool is empty, atmospheric air becomes the only heat sink. The special heat exchanger design allows to have a heat exchanger without being oversized and to have a system able to operate for unlimited period without external interventions. This innovative system provides an economic advantage as well as enhanced safety features.

  15. FOOD SAFETY CONTROL SYSTEM IN CHINA

    Institute of Scientific and Technical Information of China (English)

    Liu Wei-jun; Wei Yi-min; Han Jun; Luo Dan; Pan Jia-rong

    2007-01-01

    Most countries have expended much effort to develop food safety control systems to ensure safe food supplies within their borders. China, as one of the world's largest food producers and consumers,pays a lot of attention to food safety issues. In recent years, China has taken actions and implemented a series of plans in respect to food safety. Food safety control systems including regulatory, supervisory,and science and technology systems, have begun to be established in China. Using, as a base, an analysis of the current Chinese food safety control system as measured against international standards, this paper discusses the need for China to standardize its food safety control system. We then suggest some policies and measures to improve the Chinese food safety control system.

  16. Safety Management System in Croatia Control Ltd.

    OpenAIRE

    Pavlin, Stanislav; Sorić, Vedran; Bilać, Dragan; Dimnik, Igor; Galić, Daniel

    2009-01-01

    International Civil Aviation Organization and other international aviation organizations regulate the safety in civil aviation. In the recent years the International Civil Aviation Organization has introduced the concept of the safety management system through several documents among which the most important is the 2006 Safety Management Manual. It treats the safety management system in all the segments of civil aviation, from carriers, aerodromes and air traffic control to design, constructi...

  17. System Design and the Safety Basis

    International Nuclear Information System (INIS)

    Ellingson, Darrel

    2008-01-01

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities

  18. Safety assessment for Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Leahy, T.J.

    2012-01-01

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

  19. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

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

  20. Compositional properties of passivity

    NARCIS (Netherlands)

    Kerber, Florian; van der Schaft, Arjan

    2011-01-01

    The classical passivity theorem states that the negative feedback interconnection of passive systems is again passive. The converse statement, - passivity of the interconnected system implies passivity of the subsystems -, turns out to be equally valid. This result implies that among all feasible

  1. Software Quality Assurance for Nuclear Safety Systems

    International Nuclear Information System (INIS)

    Sparkman, D R; Lagdon, R

    2004-01-01

    The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: (sm b ullet) Ensures the software processes developed to address nuclear safety in design, operation, construction and maintenance of its facilities are safe (sm b ullet) Considers the larger system that uses the software and its impacts (sm b ullet) Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety

  2. Performance improvement of the finned passive PVT system using reflectors like removable insulation covers

    International Nuclear Information System (INIS)

    Ziapour, Behrooz M.; Palideh, Vahid; Mokhtari, Farhad

    2016-01-01

    Highlights: • A passive PVT system means the combination of a PV panel and a compact solar water heater. • Comparative study was done on performance characteristics in passive and hybrid PVT systems. • Reflectors effects on performance of a finned passive PVT system were numerically studied. • Results show that the finned passive PVT system has higher performance than the hybrid type. • Reflectors reduce the night heat losses and increase the solar radiation rate on PVT system. - Abstract: A passive photovoltaic–thermal system (PVT) is the combination of a photovoltaic (PV) panel and a compact solar water heater for co-generation of heat and electricity. This system bears considerable heat losses to ambient, particularly at noncollection times. One simple way to overcome this problem is to use a removable insulation cover on the collector's outer glazing. In this paper, the effects of the reflectors on day and night performance of a finned passive PVT system were numerically studied. At nonenergy collection time, the reflectors can turn and cover the collector cover glass as a nonconductor material. Simulation results showed that the reflectors reduce the night heat losses and increase the solar radiation rate on the absorber plate. The use of removable insulation reflectors resulted to saving extra sensibly thermal energy. Also, the solar cells power generation (P_s_c), in the case of reflectors installed, was reinforced.

  3. Autonomous system for launch vehicle range safety

    Science.gov (United States)

    Ferrell, Bob; Haley, Sam

    2001-02-01

    The Autonomous Flight Safety System (AFSS) is a launch vehicle subsystem whose ultimate goal is an autonomous capability to assure range safety (people and valuable resources), flight personnel safety, flight assets safety (recovery of valuable vehicles and cargo), and global coverage with a dramatic simplification of range infrastructure. The AFSS is capable of determining current vehicle position and predicting the impact point with respect to flight restriction zones. Additionally, it is able to discern whether or not the launch vehicle is an immediate threat to public safety, and initiate the appropriate range safety response. These features provide for a dramatic cost reduction in range operations and improved reliability of mission success. .

  4. The PIANC Safety Factor System for Breakwaters

    DEFF Research Database (Denmark)

    Burcharth, H. F.

    2000-01-01

    The paper presents a summary of the recommendations for implementation of safety in breakwater designs given by the PIANC PTC IT Working Group No 12 on Analysis of Rubble Mound Breakwaters with Vertical and Inclined Concrete Walls. The working groups developed for the most important failure modes...... a system of partial safety factors which facilitate design to any target safety level....

  5. Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

    International Nuclear Information System (INIS)

    Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

    2001-01-01

    Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

  6. PASSIVE WIRELESS MULTI-SENSOR TEMPERATURE AND PRESSURE SENSING SYSTEM USING ACOUSTIC WAVE DEVICES, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This proposal describes the development of passive surface acoustic wave (SAW) sensors and multi-sensor systems for NASA application to remote wireless sensing of...

  7. Passive Wireless Sensor System for Space and Structural Health Monitoring, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Aviana Molecular (Aviana) and the University of Central Florida (UCF) propose to develop a Passive Wireless Sensor System (PWSS) for Structural Health Monitoring...

  8. Safety-related control air systems

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This Standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This Standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this Standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements

  9. Phase Coexistence in Two-Dimensional Passive and Active Dumbbell Systems

    Science.gov (United States)

    Cugliandolo, Leticia F.; Digregorio, Pasquale; Gonnella, Giuseppe; Suma, Antonio

    2017-12-01

    We demonstrate that there is a macroscopic coexistence between regions with hexatic order and regions in the liquid or gas phase over a finite interval of packing fractions in active dumbbell systems with repulsive power-law interactions in two dimensions. In the passive limit, this interval remains finite, similar to what has been found in two-dimensional systems of hard and soft disks. We did not find discontinuous behavior upon increasing activity from the passive limit.

  10. The passive seismic aftershock Monitoring system: testing program and preliminary results

    International Nuclear Information System (INIS)

    Mokhtari, M.

    2005-01-01

    The paper is dedicated to testing program (phase of the passive seismic aftershock monitoring system with RefTek equipment (Refraction Technology, Inc., USA) for On-Site Inspection purposes that was carried out near Vienna International Centre in 2000. Equipment and applied software are described. Testing results were analyzed; in particular, least needs in maintenance personnel during operation. Development perspectives of passive seismic aftershock monitoring system for On-Site Inspection have been discussed. (author)

  11. A philosophy for space nuclear systems safety

    International Nuclear Information System (INIS)

    Marshall, A.C.

    1992-01-01

    The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions

  12. Passive ventilation systems with heat recovery and night cooling

    DEFF Research Database (Denmark)

    Hviid, Christian Anker; Svendsen, Svend

    2008-01-01

    with little energy consumption and with satisfying indoor climate. The concept is based on using passive measures like stack and wind driven ventilation, effective night cooling and low pressure loss heat recovery using two fluid coupled water-to-air heat exchangers developed at the Technical University...... simulation program ESP-r to model the heat and air flows and the results show the feasibility of the proposed ventilation concept in terms of low energy consumption and good indoor climate....

  13. Automation for System Safety Analysis

    Science.gov (United States)

    Malin, Jane T.; Fleming, Land; Throop, David; Thronesbery, Carroll; Flores, Joshua; Bennett, Ted; Wennberg, Paul

    2009-01-01

    This presentation describes work to integrate a set of tools to support early model-based analysis of failures and hazards due to system-software interactions. The tools perform and assist analysts in the following tasks: 1) extract model parts from text for architecture and safety/hazard models; 2) combine the parts with library information to develop the models for visualization and analysis; 3) perform graph analysis and simulation to identify and evaluate possible paths from hazard sources to vulnerable entities and functions, in nominal and anomalous system-software configurations and scenarios; and 4) identify resulting candidate scenarios for software integration testing. There has been significant technical progress in model extraction from Orion program text sources, architecture model derivation (components and connections) and documentation of extraction sources. Models have been derived from Internal Interface Requirements Documents (IIRDs) and FMEA documents. Linguistic text processing is used to extract model parts and relationships, and the Aerospace Ontology also aids automated model development from the extracted information. Visualizations of these models assist analysts in requirements overview and in checking consistency and completeness.

  14. Comparative performance of passive devices for piping system under seismic excitation

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Praveen, E-mail: pra_veen74@rediffmail.com [Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India); Jangid, R.S. [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai, 400076 (India); Reddy, G.R. [Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India)

    2016-03-15

    Highlights: • Correlated the analytical results obtained from the proposed analytical procedures with experimental results in the case of XPD. • Substantial reduction of the seismic response of piping system with passive devices is observed. • Significant increase in the modal damping of the piping system is noted. • There exist an optimum parameters of the passive devices. • Good amount of energy dissipation is observed by using passive devices. - Abstract: Among several passive control devices, X-plate damper, viscous damper, visco-elastic damper, tuned mass damper and multiple tuned mass dampers are popular and used to mitigate the seismic response in the 3-D piping system. In the present paper detailed studies are made to see the effectiveness of the dampers when used in 3-D piping system subjected to artificial earthquake with increasing amplitudes. The analytical results obtained using Wen's model are compared with the corresponding experimental results available which indicated a good match with the proposed analytical procedure for the X-plate dampers. It is observed that there is significant reduction in the seismic response of interest like relative displacement, acceleration and the support reaction of the piping system with passive devices. In general, the passive devices under particular optimum parameters such as stiffness and damping are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping system.

  15. Probabilistic safety criteria at the safety function/system level

    International Nuclear Information System (INIS)

    1989-09-01

    A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  16. Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. 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.

  18. Architecture Level Safety Analyses for Safety-Critical Systems

    Directory of Open Access Journals (Sweden)

    K. S. Kushal

    2017-01-01

    Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.

  19. 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)

  20. The safety interlocking system at the NAC

    International Nuclear Information System (INIS)

    Visser, K.; Mostert, H.

    1984-01-01

    The central safety interlocking system (CSIS) controls the higher level of interlocking between the various cyclotron subsystems. It ensures the safe operation of the entire cyclotron facility as regards personnel safety and proper instrument operation. The system consists of a micro-processor with a ROM-based safety interlocking program, relay output modules providing ''safety OK'' instructions to all interlocked apparatus, alarm input modules connected to transducers providing binary alarm status signals and an interface to the central control computer. All solid state electronic components of the system are situated in a low level radiation area and are interfaced to cyclotron equipment by means of 24 V relays

  1. Meeting the maglev system's safety requirements

    Energy Technology Data Exchange (ETDEWEB)

    Pierick, K

    1983-12-01

    The author shows how the safety requirements of the maglev track system derive from the general legal conditions for the safety of tracked transport. It is described how their compliance beyond the so-called ''development-accompanying'' and ''acceptance-preparatory'' safety work can be assured for the Transrapid test layout (TVE) now building in Emsland and also for later application as public transport system in Germany within the meaning of the General Railway Act.

  2. Impairment of Heat Transfer in the Passive Cooling System due to Mixed Convection

    Energy Technology Data Exchange (ETDEWEB)

    Chae Myeong Seon; Chung, Bum Jin [Kyunghee University, Yongin (Korea, Republic of); Kim, Jong Hwan [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In the passive cooling devices, the buoyant flows are induced. However the local Nusselt number of natural convective flow can be partly impaired due to the development of the mixed convective flows. This paper discusses impairment of heat transfer in the passive cooling system in relation to the development of mixed convection. The present work describes the preliminary plan to explore the phenomena experimentally. This paper is to discuss and make the plan to experiment the impairment of heat transfer in the passive cooling system due to mixed convection. In the sufficiently high passive cooling devices, the natural convection flow behavior can be mixed convection. The local Nusselt number distribution exhibits the non-monotonic behavior as axial position, since the buoyancy-aided with mixed convection was appeared. This is the part of the experimental work.

  3. The Evolution of System Safety at NASA

    Science.gov (United States)

    Dezfuli, Homayoon; Everett, Chris; Groen, Frank

    2014-01-01

    The NASA system safety framework is in the process of change, motivated by the desire to promote an objectives-driven approach to system safety that explicitly focuses system safety efforts on system-level safety performance, and serves to unify, in a purposeful manner, safety-related activities that otherwise might be done in a way that results in gaps, redundancies, or unnecessary work. An objectives-driven approach to system safety affords more flexibility to determine, on a system-specific basis, the means by which adequate safety is achieved and verified. Such flexibility and efficiency is becoming increasingly important in the face of evolving engineering modalities and acquisition models, where, for example, NASA will increasingly rely on commercial providers for transportation services to low-earth orbit. A key element of this objectives-driven approach is the use of the risk-informed safety case (RISC): a structured argument, supported by a body of evidence, that provides a compelling, comprehensible and valid case that a system is or will be adequately safe for a given application in a given environment. The RISC addresses each of the objectives defined for the system, providing a rational basis for making informed risk acceptance decisions at relevant decision points in the system life cycle.

  4. A novel passive micromixer: lamination in a planar channel system

    DEFF Research Database (Denmark)

    Tofteberg, T.; Skolimowski, Maciej; Andreassen, E.

    2010-01-01

    A novel passive micromixer concept is presented. The working principle is to make a controlled 90 degrees rotation of a flow cross-section followed by a split into several channels; the now in each of these channels is rotated a further 90 degrees before a recombination doubles the interfacial area...... between the two fluids. This process is repeated until achieving the desired degree of mixing. The rotation of the flow field is obtained by patterning the channel bed with grooves. The effect of the mixers has been Studied using computational fluid mechanics and prototypes have been micromilled in poly...

  5. Design of safety mechanism for an industrial manipulator based on passive compliance

    International Nuclear Information System (INIS)

    Kim, Hwi Su; Park, Jung Jun; Song, Jae Bok; Kyung, Jin Ho

    2010-01-01

    In recent years, collision safety between humans and robots has drawn much attention since human-robot cooperation is increasingly needed in various fields. Since positioning accuracy and collision safety are both important, an industrial manipulator should maintain very high stiffness for positioning accuracy in a normal situation, but exhibit very low stiffness when subjected to a collision force greater than the tolerance for human injury. To satisfy these requirements, we proposed in our previous research a safety mechanism composed of a linear spring and a double-slider mechanism for a service robot with a small payload. We modified this device to meet more stringent requirements for an industrial manipulator which usually has a payload higher than a service robot. Several experiments on static and dynamic collisions showed high stiffness of the safety mechanism in response to an external torque that was less than a predetermined threshold torque, but low stiffness that enabled absorption of the collision force when the external torque exceeded the threshold. Thus, positioning accuracy and collision safety were improved using the proposed design. Furthermore, a new safety criterion is suggested to verify the collision safety of a manipulator that uses the proposed safety mechanism

  6. System overview and walking dynamics of a passive dynamic walking robot with flat feet

    Directory of Open Access Journals (Sweden)

    Xinyu Liu

    2015-12-01

    Full Text Available The concept of “passive dynamic walking robot” refers to the robot that can walk down a shallow slope stably without any actuation and control which shows a limit cycle during walking. By adding actuation at some joints, the passive dynamic walking robot can walk stably on level ground and exhibit more versatile gaits than fully passive robot, namely, the “limit cycle walker.” In this article, we present the mechanical structures and control system design for a passive dynamic walking robot with series elastic actuators at hip joint and ankle joints. We built a walking model that consisted of an upper body, knee joints, and flat feet and derived its walking dynamics that involve double stance phases in a walking cycle based on virtual power principle. The instant just before impact was chosen as the start of one step to reduce the number of independent state variables. A numerical simulation was implemented by using MATLAB, in which the proposed passive dynamic walking model could walk stably down a shallow slope, which proves that the derived walking dynamics are correct. A physical passive robot prototype was built finally, and the experiment results show that by only simple control scheme the passive dynamic robot could walk stably on level ground.

  7. Appraisal of Fire Safety Management Systems at Educational Buildings

    Directory of Open Access Journals (Sweden)

    Nadzim N.

    2014-01-01

    Full Text Available Educational buildings are one type of government asset that should be protected, and they play an important role as temporary communal meeting places for children, teachers and communities. In terms of management, schools need to emphasize fire safety for their buildings. It is well known that fires are not only a threat to the building’s occupants, but also to the property and the school environment. A study on fire safety management has been carried out on schools that have recently experienced fires in Penang. From the study, it was found that the school buildings require further enhancement in terms of both active and passive fire protection systems. For instance, adequate fire extinguishers should be provided to the school and the management should inspect and maintain fire protection devices regularly. The most effective methods to increase the level of awareness on fire safety are by organizing related programs on the management of fire safety involving all staff, teachers and students, educational talks on the dangers of fire and important actions to take in the event of an emergency, and, lastly, to appoint particular staff to join the management safety team in schools.

  8. PSA methodology including new design, operational and safety factors, 'Level of recognition of phenomena with a presumed dominant influence upon operational safety' (failures of conventional as well as non-conventional passive components, dependent failures, influence of operator, fires and external threats, digital control, organizational factors)

    International Nuclear Information System (INIS)

    Jirsa, P.

    2001-10-01

    The document represents a specific type of discussion of existing methodologies for the creation and application of probabilistic safety assessment (PSA) in light of the EUR document summarizing requirements placed by Western European NPP operators on the future design of nuclear power plants. A partial goal of this discussion consists in mapping, from the PSA point of view, those selected design, operational and/or safety factors of future NPPs that may be entirely new or, at least, newly addressed. Therefore, the terms of reference for this stage were formulated as follows: Assess current level of knowledge and procedures in the analysis of factors and phenomena with a dominant influence upon operational safety of new generation reactors, especially in the following areas: (1) Phenomenology of failure types and mechanisms and reliability of conventional passive safety system components; (2) Phenomenology of failure types and mechanisms and reliability of non-conventional passive components of newly designed safety systems; (3) Phenomenology of types and mechanisms of dependent failures; (4) Human factor role in new generation reactors and its effect upon safety; (5) Fire safety and other external threats to new nuclear installations; (6) Reliability of the digital systems of the I and C system and their effect upon safety; and (7) Organizational factors in new nuclear installations. (P.A.)

  9. Analysis method for the design of a hydrogen mitigation system with passive autocatalytic recombiners in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C-H.; Sung, J-J.; Ha, S-J. [Korea Hydro and Nuclear Power Co. Ltd., Central Research Inst., Daejeon (Korea, Republic of); Yeo, I-S. [KEPCO Engineering and Construction Co. Ltd, Gyeonggi-do (Korea, Republic of)

    2014-07-01

    The importance of hydrogen safety in nuclear power plants has been emphasized especially after the Fukushima accident in Japan. A passive autocatalytic recombiner (PAR) is considered as a viable option for the mitigation of hydrogen risk because of its passive operation for hydrogen removal. This paper presents a licensed hydrogen analysis method of OPR-1000, a 1,000MWe Korea standardized pressurized water reactor with a large dry containment, to determine the capacity and locations of PARs for the design of a hydrogen mitigation system with PAR. Various accident scenarios have been adopted considering important event sequences from a combination of probabilistic methods, deterministic methods and sound engineering judgment. A MAAP 4.0.6+ with a multi-compartment model is used as an analysis tool with conservative hydrogen generation and removal models. The detailed analyses are performed for selected severe accident scenarios including sensitivity analysis with/without operations of various safety systems. The possibility of global flame acceleration (FA) and deflagration-to-detonation transient (DDT) are assessed with sigma (flame acceleration potential) and 7-lambda (DDT potential) criterion. It is concluded that the newly designed hydrogen mitigation system with twenty-four (24) PARs can effectively remove hydrogen in the containment atmosphere and prevent global FA and DDT. (author)

  10. Safety Verification for Probabilistic Hybrid Systems

    DEFF Research Database (Denmark)

    Zhang, Lijun; She, Zhikun; Ratschan, Stefan

    2010-01-01

    The interplay of random phenomena and continuous real-time control deserves increased attention for instance in wireless sensing and control applications. Safety verification for such systems thus needs to consider probabilistic variations of systems with hybrid dynamics. In safety verification o...... on a number of case studies, tackled using a prototypical implementation....

  11. Ergonomics in the context of system safety

    International Nuclear Information System (INIS)

    Donnelly, K.E.

    1984-01-01

    In a complex industrial environment, ergonomics must be combined with management science and systems analysis to produce a program which can create effective change and improve safety performance. We give an overview of such an approach, namely System Safety, so that its ergonomic content may be seen

  12. Technical self reliance of digital safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Lee, Dong Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Kook Hun [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of); Choi, Seung Gap [POSCON, Pohang (Korea, Republic of)

    2009-04-15

    This paper summarizes the development results of the Korea Nuclear Instrumentation and Control System (KNICS) project sponsored by the Korean government. In this project, Man Machine Interface System (MMIS) architecture, two digital platforms, and several control systems are developed. One platform is a programmable Logic Controller (PLC) for a safety system and another platform is a Distributed Control System (DCS) for a non safety system. With the POSAFE Q PLC, a Reactor Protection System (RPS) and an Engineered Safety Feature Component Control System (ESF CCS) are developed. A Power Control System (PCS) is developed based on the DCS. The safety grade platform and the digital safety systems obtained approval for the Topical Report from the Korean regulatory body in February of 2009. Also a Korean utility and a vendor company determined KNICS results to apply them to the planned Nuclear Power Plant (NPP) in March 2009. This paper introduces the technical self reliance experiences of the safety grade platform and the digital safety systems developed in the KNICS R and D project.

  13. Simplified analysis of passive residual heat removal systems for small size PWR's

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1992-02-01

    The function and general objectives of a passive residual heat removal system for small size PWR's are defined. The characteristic configuration, the components and the operation modes of this system are concisely described. A preliminary conceptual specification of this system, for a small size PWR of 400 MW thermal, is made analogous to the decay heat removal system of the AP-600 reactor. It is shown by analytic models that such passive systems can dissipate 2% of nominal power within the thermal limits allowed to the reactor fuel elements. (author)

  14. Systems Thinking and Patient Safety

    National Research Council Canada - National Science Library

    Schyve, Paul M

    2005-01-01

    Patient safety is a prominent theme in health care delivery today. This should come as no surprise, given that "first, do no harm" has been the ethical watchword throughout the history of medicine, nursing, and pharmacy...

  15. A practical approach to harmonic compensation in power systems-series connection of passive and active filters

    OpenAIRE

    Fujita, Hideaki; Akagi, Hirofumi

    1991-01-01

    The authors present a combined system with a passive filter and a small-rated active filter, both connected in series with each other. The passive filter removes load produced harmonics just as a conventional filter does. The active filter plays a role in improving the filtering characteristics of the passive filter. This results in a great reduction of the required rating of the active filter and in eliminating all the limitations faced by using only the passive filter, leading to a practica...

  16. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  17. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

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

    International Nuclear Information System (INIS)

    Lee, Hyung Sok

    2012-02-01

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

  19. Safety-critical Java for embedded systems

    DEFF Research Database (Denmark)

    Schoeberl, Martin; Dalsgaard, Andreas Engelbredt; Hansen, René Rydhof

    2016-01-01

    This paper presents the motivation for and outcomes of an engineering research project on certifiable Javafor embedded systems. The project supports the upcoming standard for safety-critical Java, which defines asubset of Java and libraries aiming for development of high criticality systems....... The outcome of this projectinclude prototype safety-critical Java implementations, a time-predictable Java processor, analysis tools formemory safety, and example applications to explore the usability of safety-critical Java for this applicationarea. The text summarizes developments and key contributions...

  20. Operation safety of complex industrial systems

    International Nuclear Information System (INIS)

    Zwingelstein, G.

    1999-01-01

    Zero fault or zero risk is an unreachable goal in industrial activities like nuclear activities. However, methods and techniques exist to reduce the risks to the lowest possible and acceptable level. The operation safety consists in the recognition, evaluation, prediction, measurement and mastery of technological and human faults. This paper analyses each of these points successively: 1 - evolution of operation safety; 2 - definitions and basic concepts: failure, missions and functions of a system and of its components, basic concepts and operation safety; 3 - forecasting analysis of operation safety: reliability data, data-banks, precautions for the use of experience feedback data; realization of an operation safety study: management of operation safety, quality assurance, critical review and audit of operation safety studies; 6 - conclusions. (J.S.)

  1. A passive decay-heat removal system for an ABWR based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2017-01-15

    Highlights: • A passive decay heat removal system for an ABWR is discussed using combined system of the reactor and an air cooler. • Effect of number of pass of the finned heat transfer tubes on heat removal is investigated. • The decay heat can be removed by air coolers with natural convection. • Two types of air cooler are evaluated, i.e., steam condensing and water cooling types. • Measures how to improve the heat removal rate and to make compact air cooler are discussed. - Abstract: This paper describes the capability of an air cooling system (ACS) operated under natural convection conditions to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA). The plant suffered damages due to the tsunami and entered a state of Station Blackout (SBO) during which seawater cooling was not available. To prevent this kind of situation, we proposed a passive decay heat removal system (DHRS) in the previous study. The plant behavior during the SBO was calculated using the system code NETFLOW++ assuming an ABWR with the ACS. However, decay heat removal under an air natural convection was difficult. In the present study, a countermeasure to increase heat removal rate is proposed and plant transients with the ACS are calculated under natural convection conditions. The key issue is decreasing pressure drop over the tube banks in order to increase air flow rate. The results of the calculations indicate that the decay heat can be removed by the air natural convection after safety relief valves are actuated many times during a day. Duct height and heat transfer tube arrangement of the AC are discussed in order to design a compact and efficient AC for the natural convection mode. As a result, a 4-pass heat transfer tubes with 2-row staggered arrangement is the candidate of the AC for the DHRS under the air natural convection conditions. The heat removal rate is re-evaluated as

  2. Molten salt reactor as asymptotic safety nuclear system

    International Nuclear Information System (INIS)

    Novikov, V.M.; Ignatyev, V.V.

    1989-01-01

    Safety is becoming the main and priority problem of the nuclear power development. An increase of the active safety measures could hardly be considered as the proper way to achieve the asymptotically high level of nuclear safety. It seem that the more realistic way to achieve such a goal is to minimize risk factors and to maximize the use of inherent and passive safety properties. The passive inherent safety features of the liquid fuel molten salt reactor (MSR) technology are making it attractive for future energy generation. The achievement of the asymptotic safety in MSR is being connected with the minimization of such risk factors as a reactivity excess, radioactivity stored, decay heat, non nuclear energy stored in core. In this paper safety peculiarities of the different MSR concepts are discussed

  3. Overview of Indian position: Passive and active safety features of LMFBRs

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1991-01-01

    Since PWR and BWR type reactors are considered to have acceptable level of safety, there is a general statement that LMFBR type reactors should be designed to be 'at least as safe' as PWRs and BWRs. Nuclear safety issues considered include the following: flexibility of operation over the entire power range, effectiveness of steps ensuring the safety as well as plant availability under foreseeable incident conditions; guaranteed safe shut-down of the reactor; ability to remove safely and reliably the decay heat under shut-down conditions; containment capability including protection of the reactor from external events. It is concluded that liquid metal fast breeder reactors (LMFBRs) can be designed very easily to eliminate anxieties about their safety and no difficulties are expected in the licensing procedure of this type of reactors

  4. The Rush to Remediate: Long Term Performance Favors Passive Systems at SRS

    International Nuclear Information System (INIS)

    Hoffman, D.; Cauthen, K.; Beul, R. R.

    2003-01-01

    The purpose of this paper is to describe the long-term performance of groundwater remediation systems at SRS and compare active versus passive systems. The presentation will focus on the limited effectiveness of active pump and treat systems and share the experience with more passive and natural systems such as soil vapor extraction, barometric pumping, bioremediation, and phytoremediation. Three remediation projects are presented. In each case the waste source is capped with clay or synthetic barriers; however, extensive groundwater contamination remains. The first project features the cleanup of the largest plume in the United States. The second project entails solvent and vinyl chloride remediation of groundwater beneath a hazardous waste landfill. The third project discusses tritium containment from a 160-acre radioactive waste disposal area. Special emphasis is placed on performance data from alternate technology cleanup. The goals are to share remediation data, successes and lessons learned, while making a case for passive systems use in groundwater remediation

  5. 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

  6. 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.