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Sample records for engineered safety features

  1. Technical specification optimization program - engineered safety features

    International Nuclear Information System (INIS)

    Andre, G.R.; Jansen, R.L.

    1986-01-01

    The Westinghouse Technical Specification Program (TOP) was designed to evaluate on a quantitative basis revisions to Nuclear Power Plant Technical Specifications. The revisions are directed at simplifying plant operation, and reducing unnecessary transients, shutdowns, and manpower requirements. In conjunction with the Westinghouse Owners Group, Westinghouse initiated a program to develop a methodology to justify Technical Specification revisions; particularly revisions related to testing and maintenance requirements on plant operation for instrumentation systems. The methodology was originally developed and applied to the reactor trip features of the reactor protection system (RPS). The current study further refined the methodology and applied it to the engineered safety features of the RPS

  2. Definitions of engineered safety features and related features for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    In light water moderated, light water cooled nuclear power plants, definitions are given of engineered safety features which are designed to suppress or prevent dispersion of radioactive materials due to damage etc. of fuel at the times of power plant failures, and of related features which are designed to actuate or operate the engineered safety features. Contents are the following: scope of engineered safety features and of related features; classification of engineered safety features (direct systems and indirect systems) and of related features (auxiliaries, emergency power supply, and protective means). (Mori, K.)

  3. Engineering safety features for high power experimental reactors

    International Nuclear Information System (INIS)

    Doval, A.; Villarino, E.; Vertullo, A.

    2000-01-01

    In the present analysis we will focus our attention in the way engineering safety features are designed in order to prevent fuel damage in case of abnormal or accidental situations. To prevent fuel damage two main facts must be considered, the shutdown of the reactor and the adequate core cooling capacity, it means that both, neutronic and thermohydraulic aspects must be analysed. Some neutronic safety features are common to all power ranges like negative feedback reactivity coefficients and the required number of control rods containing the proper absorber material to shutdown the reactor. From the thermohydraulic point of view common features are siphon-breaker devices and flap valves for those powers requiring cooling in the forced convection regime. For the high power reactor group, the engineering safety features specially designed for a generic reactor of 20 MW, will be presented here. From the neutronic point of view besides the common features, and to comply with our National Regulatory Authority, a Second Shutdown System was designed as a redundant shutdown system in case the control plates fail. Concerning thermohydraulic aspects besides the pump flywheels and the flap valves providing the natural convection loop, a metallic Chimney and a Chimney Water Injection System were supplied. (author)

  4. USNRC regulatory guidance for engineered safety feature air cleaning systems

    International Nuclear Information System (INIS)

    Bellamy, R.R.

    1991-01-01

    The need for clear, technically appropriate, and easily implementable guidance for the design, testing, and maintenance of nuclear air cleaning systems has long been recognized. Numerous industry consensus standards have been issued and revised over the last 30 years. Guidance has also been published by the US Nuclear Regulatory Commission in the form of regulations, regulatory guides, standard review plans, NUREG documents, and information notices. This paper will summarize the latest revisions to these documents and emphasize Regulatory Guide 1.52, Design, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, which was last revised in 1978. The USNRC has undertaken a project to revise this regulatory guide, and the status of that revision is highlighted

  5. Engineering and safety features of modular vault dry storage

    International Nuclear Information System (INIS)

    Deacon, D.; Wheeler, D.J.

    1984-01-01

    This paper discusses the need for interim dry storage and reviews detailed features of the Modular Vault Dry storage concept. The concept meets three basic utility requirements. Firstly, the technology and safety features have been demonstrated on existing plant; secondly, it can be built and licensed in an acceptably short timescale; and thirdly, economic analysis shows that a modular vault dry store is often the cheapest option for interim storage

  6. New source terms: what do they tell us about engineered safety feature performance

    International Nuclear Information System (INIS)

    Bernero, R.M.

    1985-01-01

    The accident behavior models which are the basis of engineered safety feature design are generally simple, non-mechanistic and concentrated on volatile radioiodine. Now data from source term studies show that models should be more mechanistic and look at other species than volatile iodine. A complete reevaluation of engineered safety features is needed

  7. Fission product source terms and engineered safety features

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1984-01-01

    The author states that new, technically defensible, methodologies to establish realistic source term values for nuclear reactor accidents will soon be available. Although these methodologies will undoubtedly find widespread use in the development of accident response procedures, the author states that it is less clear that the industry is preparing to employ the newer results to develop a more rational approach to strategies for the mitigation of fission product releases. Questions concerning the performance of existing engineered safety systems are reviewed

  8. 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

  9. What price safety. A probabilistic cost-benefit evaluaton of existing engineered safety features

    International Nuclear Information System (INIS)

    O'Donnell, E.P.

    1978-01-01

    The paper provides a method for performing quantitative cost-benefit evaluations for nuclear safety concerns involving accidents of low probability and potentially large consequences. It presents an application of the method to ECCS, containment, emergency power system and hydrogen recombiner system. This evaluation provides a valuable assessment of the relative cost effectiveness of these features in reducing accident risk. It also provides insight into the sensitivity of cost-benefit calculations to the manner in which safety features are sequantially added in design. (author)

  10. Technical considerations for the development of an engineering safety features control system with PLC

    International Nuclear Information System (INIS)

    Lee, C. K.; Kim, C. H.; Han, J. B.; Kim, H.; Lee, S. S.

    2002-01-01

    Technical considerations are summarized for the development of an ESFCS(Engineered Safety Features Control System) with PLC (Programmable Logic Controller). The ESFCS is required for the mitigation of plant accident conditions and therefore developed in conformance with the design requirements applied to the safety critical system. The design of ESFCS primarily considered its safety, and the system has an architecture that will be able to minimize spurious actuation. The PLC based functional distribution and redundant design features are adopted, and the fieldbus is applied in the communication of information and control signals between PLC processors. It is expected that the ESFCS will have several advanced design features compared with the conventional systems supplied by foreign vendors

  11. Design of the Control System for Engineered Safety Features of KIJANG Research Reactor

    International Nuclear Information System (INIS)

    Kim, Hagtae; Kim, Jun-Yeon; Chae, Hee-Taek

    2015-01-01

    The purpose of this paper is to design an effective control system for the Engineered Safety Features (ESF) of KJRR such as the Safety Residual Heat Removal System (SRHRS) pumps and Siphon Break Valve (SBV) without an Engineered Safety Features-Component Control System (ESF-CCS). This control system is called a 'local motor starter', because this system controls motors in the SRHRS pumps and SBVs by receiving the signal from Reactor Protection System (RPS) and Alternate Protection System (APS) when the differential pressure or pool level reach the set points. In this paper, the design concepts and requirements of the local motor starter based on the design features of KJRR is proposed. An ESF is a safety system that mitigates consequences of the Anticipated Operational Occurrence (AOO) and Design Basis Accident (DBA). The results of this paper are able to be used for the development of control systems for research reactors similar to KJRR. The precondition for such application is to have a few ESFs and conduct simple logic. The proposed control system called a local motor starter is being designed, and a manufacture of the actual systems is expected in the foreseeable future

  12. An Axiomatic Design Approach of Nanofluid-Engineered Nuclear Safety Features for Generation III+ React

    International Nuclear Information System (INIS)

    Bang, In Cheol; Heo, Gyun Young; Jeong, Yong Hoon; Heo, Sun

    2009-01-01

    A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems

  13. Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

    1984-01-01

    The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

  14. Ageing study of the engineered safety features actuation system of the Loviisa NPP

    International Nuclear Information System (INIS)

    Simola, K.; Maskuniitty, M.

    1995-06-01

    An ageing study of the engineered safety features actuation system of the Loviisa nuclear power plant has been performed. The operating experience, including failure and maintenance histories of analog measuring devices, logics for safety signal formation and individual control electronics of pumps and valves, has been collected and analysed. The safety importance of system components has been studied with a fault tree analysis of a selected safety function. Based on the results of the analysis of operating experiences and the fault tree analysis, some components were selected for deeper analyses. According to the operating experience, the amount of failures in the Loviisa plant safety system has been low and no increasing trend in the failure history can yet be observed. Only a few failures had prohibited the propagation of the safety signal, mostly the failures have caused a false alarm. The failures reported have concerned mainly limit signal units, transmitters, and priority units. According to the fault tree analysis of one safety function, the most important components of this subsystem are individual control units and pulse/DC converters. Failure modes and effect analyses were performed for priority and individual control unit, limit signal unit and comparator and pulse/DC converter in order to identify the critical failure modes of these devices. (orig.) (15 refs., 26 figs., 9 tabs.)

  15. New design of engineered safety features-component control system to improve performance and reliability

    International Nuclear Information System (INIS)

    Kim, S.T.; Jung, H.W.; Lee, S.J.; Cho, C.H.; Kim, D.H.; Kim, H.

    2006-01-01

    Full text: Full text: The Engineered Safety Features-Component Control System (ESF-CCS) controls the engineered safety features of a Nuclear Power Plant such as Solenoid Operated Valves (SOV), Motor Operated Valves (MOV), pumps, dampers, etc. to mitigate the effects of a Design Basis Accident (DBA) or an abnormal operation. ESF-CCS serves as an interface system between the Plant Protection System (PPS) and remote actuation devices. ESF-CCS is composed of fault tolerant Group Controllers GC, Loop Controllers (LC), ESF-CCS Test and Interface Processor (ETIP) and Cabinet Operator Module (COM) and Control Channel Gateway (CCG) etc. GCs in each division are designed to be fully independent triple configuration, which perform system level NSSS and BOP ESFAS logic (2-out-of-4 logic and l-out-of-2 logic, respectively) making it possible to test each GC individually during normal operation. In the existing configuration, the safety-related plant component control is part of the Plant Control System (PCS) non-safety system. For increased safety and reliability, this design change incorporates this part into the LCs, and is therefore designed according to the safety-critical system procedures. The test and diagnosis capabilities of ETIP and COM are reinforced. By means of an automatic periodic test for all main functions of the system, it is possible to quickly determine an abnormal status of the system, and to decrease the elapsed time for tests, thus effectively increasing availability. ESF-CCS consists of four independent divisions (A, B, C, and D) in the Advanced Power Reactor 1400 (APR1400). One prototype division is being manufactured and will be tested

  16. Implementation of special engineering safety features for severe accident management. New SAMG approach

    International Nuclear Information System (INIS)

    Grigorov, D.; Borisov, E.; Mancheva, K.

    2012-01-01

    Conclusions: As a result of the thermohydraulic analysis conducted the following main conclusions are formulated: The operator actions for accident management are effective and allow reaching conditions for application of the new engineering safety features for SAMG; The new engineering safety features application is effective and prevents severe core damage for Scenario 1. For the Scenario 2 they prevents degradation and relocation of the reactor core for a long period of time (in the analysis this period is 10 h, but the unit could be kept in safe condition for longer time which is not specifically analysed).The maximal fuel cladding temperature for Scenario 1 reaches 558 o C. This low fuel cladding temperature gradient is achieved by applying a complex of operator actions which prevent any core damage. If the additional discharge line with DN 100 mm from the PRZ is not opened then a severe core damage occurs; The maximal fuel cladding temperature for Scenario 2 reaches 1307 o C. One of the possibilities for keeping this temperature below 1200 o C is to mount second line (the first SFP line is between YT12S03.S04) from the SFP to the TQ22 pipeline which is connected to YT14B01 hydroaccumulator line, between the check valves YT14S03.S04

  17. Reliability study: digital engineered safety feature actuation system of Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sudarno; Kang, H. G.; Jang, S. C.; Eom, H. S.; Ha, J. J.

    2003-04-01

    The usage of digital Instrumentation and Control (I and C) in a nuclear power plant becomes more extensive, including safety related systems. The PSA application of these new designs are very important in order to evaluate their reliability. In particular, Korean Standard Nuclear Power Plants (KSNPPs), typically Ulchin 5 and 6 (UCN 5 and 6) reactor units, adopted the digital safety-critical systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS). In this research, we developed fault tree models for assessing the unavailability of the DESFAS functions. We also performed an analysis of the quantification results. The unavailability results of different DESFAS functions showed that their values are comprised from 5.461E-5 to 3.14E-4. The system unavailability of DESFAS AFAS-1 is estimated as 5.461E-5, which is about 27% less than that of analog system if we consider the difference of human failure probability estimation between both analyses. The results of this study could be utilized in risk-effect analysis of KSNPP. We expect that the safety analysis result will contribute to design feedback

  18. 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.

  19. Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

    International Nuclear Information System (INIS)

    Lee, Sangseok; Sohn, Kwangyoung; Lee, Junku; Park, Geunok

    2013-01-01

    Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of controllers

  20. Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sangseok; Sohn, Kwangyoung [Korea Reliability Technology and System, Daejeon (Korea, Republic of); Lee, Junku; Park, Geunok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of

  1. Gaseous core nuclear-driven engines featuring a self-shutoff mechanism to provide nuclear safety

    International Nuclear Information System (INIS)

    Heidrich, J.; Pettibone, J.; Chow, Tze-Show; Condit, R.; Zimmerman, G.

    1991-11-01

    Nuclear driven engines are described that could be run in either pulsed or steady state modes. In the pulsed mode nuclear energy is released by fissioning of uranium or plutonium in a supercritical assembly of fuel and working gas. In a steady state mode a fuel-gas mixture is injected into a magnetic nozzle where it is compressed into a critical state and produces energy. Engine performance is modeled using a code that calculates hydrodynamics, fission energy production, and neutron transport self-consistently. Results are given demonstrating a large negative temperature coefficient that produces self-shutoff or control of energy production. Reduced fission product inventory and the self-shutoff provide inherent nuclear safety. It is expected that nuclear engine reactor units could be scaled up from about 100 MW e

  2. The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

    International Nuclear Information System (INIS)

    Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

    2007-01-01

    The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods

  3. FISSION 2120: a program for assessing the need for engineered safety feature grade air cleaning systems in post accident environments

    International Nuclear Information System (INIS)

    Martin, G. Jr.; Michlewicz, D.; Thomas, J.

    1979-01-01

    A computer program FISSION 2120, has been developed to evaluate the need for various engineered Safety Feature grade air cleaning systems to mitigate radiation exposures resulting from accidential releases of radioactivity. Those systems which are generally investigated include containment sprays with chemical additives, containment fan coolers with charcoal filters, and negative pressure maintenance systems for double barrier containments with either one-pass filtration or recirculation with filtration. The program can also be used to calculate the radiation doses to control room personnel. This type of analysis is directed towards the various protection aspects of the emergency ventilation system and involves the modeling of the radiological source terms and the atmospheric transport of the radioactive releases. The modeling is enhanced by the inherent capability of the program to accommodate simultaneous release of activity from several sources and to perform a dose evaluation for a wide range of the design characteristics of control room emergency air filtration systems. Use of the program has resulted in considerable savings in the time required to perform such analyses and in the selection of the most cost-effective Engineered Safety Features

  4. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  5. Engineering features of ISX

    International Nuclear Information System (INIS)

    Lousteau, D.C.; Jernigan, T.C.; Schaffer, M.J.; Hussung, R.O.

    1975-01-01

    ISX, an Impurity Study Experiment, is presently being designed at Oak Ridge National Laboratory as a joint scientific effort between ORNL and General Atomic Company. ISX is a moderate size tokamak dedicated to the study of impurity production, diffusion, and control. The significant engineering features of this device are discussed

  6. Suitability review of FMEA and reliability analysis for digital plant protection system and digital engineered safety features actuation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, I. S.; Kim, T. K.; Kim, M. C.; Kim, B. S.; Hwang, S. W.; Ryu, K. C. [Hanyang Univ., Seoul (Korea, Republic of)

    2000-11-15

    Of the many items that should be checked out during a review stage of the licensing application for the I and C system of Ulchin 5 and 6 units, this report relates to a suitability review of the reliability analysis of Digital Plant Protection System (DPPS) and Digital Engineered Safety Features Actuation System (DESFAS). In the reliability analysis performed by the system designer, ABB-CE, fault tree analysis was used as the main methods along with Failure Modes and Effect Analysis (FMEA). However, the present regulatory technique dose not allow the system reliability analysis and its results to be appropriately evaluated. Hence, this study was carried out focusing on the following four items ; development of general review items by which to check the validity of a reliability analysis, and the subsequent review of suitability of the reliability analysis for Ulchin 5 and 6 DPPS and DESFAS L development of detailed review items by which to check the validity of an FMEA, and the subsequent review of suitability of the FMEA for Ulchin 5 and 6 DPPS and DESFAS ; development of detailed review items by which to check the validity of a fault tree analysis, and the subsequent review of suitability of the fault tree for Ulchin 5 and 6 DPPS and DESFAS ; an integrated review of the safety and reliability of the Ulchin 5 and 6 DPPS and DESFAS based on the results of the various reviews above and also of a reliability comparison between the digital systems and the comparable analog systems, i.e., and analog Plant Protection System (PPS) and and analog Engineered Safety Features Actuation System (ESFAS). According to the review mentioned above, the reliability analysis of Ulchin 5 and 6 DPPS and DESFAS generally satisfies the review requirements. However, some shortcomings of the analysis were identified in our review such that the assumed test periods for several equipment were not properly incorporated in the analysis, and failures of some equipment were not included in the

  7. Effect of engineered safety features on the risk of hypothetical LMFBR accidents

    International Nuclear Information System (INIS)

    Cybulskis, P.

    1978-01-01

    The risks of hypothetical core-disruptive accidents in liquid-metal-cooled fast breeder reactors which involve meltthrough of the reactor vessel are compared for two plant designs: one design without specific provisions to accommodate such an accident and the other design with an ex-vessel core catcher and a cvity hot liner. The approach to risk analysis used is that developed in the Reactor Safety Study (WASH-1400). Since the probability of occurrence of such an event has not been evaluated, however, insight into the potential risk is gained only on a relative basis. The principal conclusions of this study are: (1) adding a core catcher--hot liner reduces the probabilty of accidents having major consequences; (2) the degree to which hot liner--core catcher systems can reduce the risk of melt-through accidents is limited by the failure probability of these systems; (3) fractional radioactive releases to the environment in the liquid-metal-cooled fast breeder reactor accidents considered are comparable to those from the light-water reactors evaluated in WASH-1400; (4) since sodium--concrete reactions are a dominant driving force during the accident, the integrity of the cavity liner is as important as the function of the core catcher; (5) there may be other accidents or paths to radioactive releases that are not affected by the addition of a hot liner--core catcher

  8. Reliability and safety engineering

    CERN Document Server

    Verma, Ajit Kumar; Karanki, Durga Rao

    2016-01-01

    Reliability and safety are core issues that must be addressed throughout the life cycle of engineering systems. Reliability and Safety Engineering presents an overview of the basic concepts, together with simple and practical illustrations. The authors present reliability terminology in various engineering fields, viz.,electronics engineering, software engineering, mechanical engineering, structural engineering and power systems engineering. The book describes the latest applications in the area of probabilistic safety assessment, such as technical specification optimization, risk monitoring and risk informed in-service inspection. Reliability and safety studies must, inevitably, deal with uncertainty, so the book includes uncertainty propagation methods: Monte Carlo simulation, fuzzy arithmetic, Dempster-Shafer theory and probability bounds. Reliability and Safety Engineering also highlights advances in system reliability and safety assessment including dynamic system modeling and uncertainty management. Cas...

  9. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    International Nuclear Information System (INIS)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-01-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs

  10. 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...

  11. FEATURES ROAD SAFETY AUDIT

    Directory of Open Access Journals (Sweden)

    L. Abramova

    2015-07-01

    Full Text Available Development of the road network, increasing motorization of the population significantly increase the risk of accidents. Experts in the field of traffic are developing methods to reduce the probability of accidents. The ways of solving the problems of road safety audit at various stages of the «life» of roads are considered.

  12. Environmental, safety, and health engineering

    International Nuclear Information System (INIS)

    Woodside, G.; Kocurek, D.

    1997-01-01

    A complete guide to environmental, safety, and health engineering, including an overview of EPA and OSHA regulations; principles of environmental engineering, including pollution prevention, waste and wastewater treatment and disposal, environmental statistics, air emissions and abatement engineering, and hazardous waste storage and containment; principles of safety engineering, including safety management, equipment safety, fire and life safety, process and system safety, confined space safety, and construction safety; and principles of industrial hygiene/occupational health engineering including chemical hazard assessment, personal protective equipment, industrial ventilation, ionizing and nonionizing radiation, noise, and ergonomics

  13. Study of long-term loss of all AC power supply sources for VVER-1000/V320 in connection with application of new engineering safety features for SAMG

    International Nuclear Information System (INIS)

    Borisov, Evgeni; Grigorov, Dobrin; Mancheva, Kaliopa

    2013-01-01

    Highlights: • In this study we presented analysis for a new SAMG approach. • The approach is applicable for all PWR reactors from 2nd generation. • We investigated two scenarios with total black out. • The RELAP/MOD 3.2 computer code is used in performing the analyses. - Abstract: This paper presents the results of analysis for application of a new Severe Accident Management Guideline (SAMG) approach which is specifically applied for VVER-1000/B320 reactor installations. In general, this innovative approach is fully applicable for all the pressurized water reactors from second generation. The purposes of the analysis for the new SAMG approach application are as follows: • To represent suggestions for new engineering safety features application for SAMG strategies. • To assess the applicability of the new engineering safety features and means for SAMG strategies in case of loss of all off-site power supply sources for VVER-1000/B320 reactor installations. • To represent important operator actions and to analyse the effectiveness of these actions for accidents management in compliance with the new approach. • The RELAP5/MOD3.3 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. The input data deck for the analysis is optimized, verified and validated

  14. Fire safety engineering

    International Nuclear Information System (INIS)

    Smith, D.N.

    1989-01-01

    The periodic occurrence of large-scale, potentially disastrous industrial accidents involving fire in hazardous environments such as oilwell blowouts, petrochemical explosions and nuclear installations highlights the need for an integrated approach to fire safety engineering. Risk reduction 'by design' and rapid response are of equal importance in the saving of life and property in such situations. This volume of papers covers the subject thoroughly, touching on such topics as hazard analysis, safety design and testing, fire detection and control, and includes studies of fire hazard in the context of environment protection. (author)

  15. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  16. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: a pool-type primary system, and advanced ternary alloy metallic fuel, and an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  17. The Safety Feature Test of QNX RTOS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jang Yeol; Lee, Young Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Benchmarking is a point of reference by which something can be measured. The QNX is a kind of Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available PC to be applied such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS-AC). The ELMSYS PC Hardware will be qualified by KTL(Korea Testing Lab.) in order to use as a Cabinet Operator Module(COM). QNX RTOS is dedicating by KAERI now. This paper describes the outline and some safety features among benchmarking test for QNX RTOS under the ELMSYS PC platform

  18. The Safety Feature Test of QNX RTOS

    International Nuclear Information System (INIS)

    Kim, Jang Yeol; Lee, Young Jun

    2010-01-01

    Benchmarking is a point of reference by which something can be measured. The QNX is a kind of Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available PC to be applied such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS-AC). The ELMSYS PC Hardware will be qualified by KTL(Korea Testing Lab.) in order to use as a Cabinet Operator Module(COM). QNX RTOS is dedicating by KAERI now. This paper describes the outline and some safety features among benchmarking test for QNX RTOS under the ELMSYS PC platform

  19. Engineered-safety-feature air-cleaning systems for commercial light-water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Burchsted, C.A.

    1975-01-01

    Substantial improvement has been observed in the design and construction of ESF air cleaning systems in some of the newer power plants, as compared to earlier practice, but there is still much to be done. Adequate space must be provided for these facilities in the earliest containment and building layout, and system designers, equipment designers, and building layout engineers must give adequate consideration to easy access to facilitate maintenance and testing. Finally, constructors and utilities must provide for proper storage of critical components such as HEPA filters and adsorber cells during construction and during the period awaiting startup of the plant. (U.S.)

  20. Mental Workload and Situational Awareness Evaluation of APR1400 Engineered Safety Features- Component Control Activation Systems using Augmented Reality

    International Nuclear Information System (INIS)

    Murungi, Mwongeera; Jung, JaeCheon

    2016-01-01

    In the study, an Augmented Reality procedure guidance support system concept was designed and used as a tool for the measurement of mental workload and Situational awareness of an SRO (Senior Reactor Operator). The EOP was chosen as the scenario for testing because it is the one of the critical plant conditions that requires human intervention and it represents (one of the more) conservative approaches to the test scenarios that are possible. The system is expected to realize an improvement in the level of Situational Awareness and mental workload which have been demonstrated by previous studies to be directly linked with the system response to an emergency situation in the MCR. The planning and design of the project adhered to a Systems Engineering approach in order to provide an optimized framework for ensuring the successful implementation of the system design. Previous study and research into this topic has emphasized the importance of situational awareness in determining the human factor performance issues in the nuclear power plant Control Room operations. This paper broadly defined a technique that successfully used the operator’s mental workload (using NASATLX) and Situational Awareness (using SART) as quantifying measures to evaluate the performance of specific ESF-CCS functions based on human factors. These results show that an improvement of the SA/workload could lead to an improvement of the level of certainty that the emergency situation can be brought under control. It is expected that future development work in this area will yield an actualized Augmented Reality system that could incorporate MCR team control and possibly be implemented in the system validation of other I and C systems

  1. Mental Workload and Situational Awareness Evaluation of APR1400 Engineered Safety Features- Component Control Activation Systems using Augmented Reality

    Energy Technology Data Exchange (ETDEWEB)

    Murungi, Mwongeera; Jung, JaeCheon [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    In the study, an Augmented Reality procedure guidance support system concept was designed and used as a tool for the measurement of mental workload and Situational awareness of an SRO (Senior Reactor Operator). The EOP was chosen as the scenario for testing because it is the one of the critical plant conditions that requires human intervention and it represents (one of the more) conservative approaches to the test scenarios that are possible. The system is expected to realize an improvement in the level of Situational Awareness and mental workload which have been demonstrated by previous studies to be directly linked with the system response to an emergency situation in the MCR. The planning and design of the project adhered to a Systems Engineering approach in order to provide an optimized framework for ensuring the successful implementation of the system design. Previous study and research into this topic has emphasized the importance of situational awareness in determining the human factor performance issues in the nuclear power plant Control Room operations. This paper broadly defined a technique that successfully used the operator’s mental workload (using NASATLX) and Situational Awareness (using SART) as quantifying measures to evaluate the performance of specific ESF-CCS functions based on human factors. These results show that an improvement of the SA/workload could lead to an improvement of the level of certainty that the emergency situation can be brought under control. It is expected that future development work in this area will yield an actualized Augmented Reality system that could incorporate MCR team control and possibly be implemented in the system validation of other I and C systems.

  2. Safety features and licensing of CNNC-ACP100

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, F., E-mail: Zhongfj2000@163.com [Nuclear Power Inst. of China, National Key Lab. of Science and Technology on Reactor System Design Technology (China)

    2014-07-01

    ACP100 is an innovatory modular pressurized water reactor, the engineering safety systems fully adopt passive safety design technology. Its inherent safety and passive features/systems are verified via testing facilities and are highlighted at certain levels of defence in depth. The licensing of ACP 100 is within current LWR framework and meets up-to-date codes and requirements in nuclear safety. (author)

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

    International Nuclear Information System (INIS)

    Hadid Subki, M.; )

    2014-01-01

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

  4. 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.

  5. Safety engineering with COTS components

    International Nuclear Information System (INIS)

    O'Halloran, Mark; Hall, Jon G.; Rapanotti, Lucia

    2017-01-01

    Safety-critical systems are becoming more widespread, complex and reliant on software. Increasingly they are engineered through (COTS) (Commercial Off The Shelf) components to alleviate the spiralling costs and development time, often in the context of complex supply chains. A parallel increased concern for safety has resulted in a variety of safety standards, with a growing consensus that a safety life cycle is needed which is fully integrated with the design and development life cycle, to ensure that safety has appropriate influence on the design decisions as system development progresses. In this article we explore the application of an integrated approach to safety engineering in which assurance drives the engineering process. The paper reports on the outcome of a case study on a live industrial project with a view to evaluate: its suitability for application in a real-world safety engineering setting; its benefits and limitations in counteracting some of the difficulties of safety engineering with (COTS) components across supply chains; and, its effectiveness in generating evidence which can contribute directly to the construction of safety cases. - Highlights: • Assurance as effective driver for COTS-based safety-critical system development. • Engages stakeholders, captures requirements and provides rich traceability. • Shares appropriate safety requirements across the supply chain.

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

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

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

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

    International Nuclear Information System (INIS)

    Hackett, D.B.

    1980-01-01

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

  8. Technical features of ABWR safety systems

    International Nuclear Information System (INIS)

    Sugisaki, Toshihiko; Tominaga, Kenji; Horiuchi, Tetsuo

    1986-01-01

    The engineering safety facilities of ABWRs have been disigned so as to have many excellent characteristics such as safety, reliability and economy, reflecting the merit of adopting new technology such as internal pumps and new control rod driving mechanism, and coupled with the safety peculiar to BWRs. In this paper, about ECCS, containment vessels and others which compose the engineering safety facilities of ABWRs, the characteristics related to the safety owing to the adoption of internal pumps and others, and the evaluation of the performance at the time of various accidents are discussed. As the results of safety evaluation, it was clarified that due to the safety peculiar to ABWRs and the characteristics of the safety facilities, the large increases of safety, reliability and economy have been planned in the ABWRs, and for example, core flooding can be maintained even at the time of a hypothetical loss of coolant accident. BWRs have the simple system constitution, good self controllability, large natural circulation ability, simple operation control method and excellent ability of confining heat and radioactivity. BWRs have three safety functions to stop reactors, to remove heat from reactors, and to confine radioactive substances. These functions of ABWRs were evaluated, and very high safety was confirmed. (Kako, I.)

  9. Improvement of the abnormal diagnosis technology by the development of an abnormal parts assignment system for the engineered safety features actuating system of the HTTR

    International Nuclear Information System (INIS)

    Hirato, Yoji; Kozawa, Takayuki; Saito, Kenji

    2015-01-01

    The safety protection sequence panel of HTTR is a control panel to actuate an engineering safety system for protecting the reactor core, reactor coolant pressure boundary, and containment vessel boundary at the time of an accident of the nuclear reactor facilities. The safety code stipulates that the control panel should receive safety check at a frequency of once a month during reactor operation. When abnormality has been found, it is required to eliminate its causes and restore normal operation as soon as possible. However, since this control panel is composed of a complex control circuit, the cause check during abnormality requires the confirmation by a knowledgeable person spending quite a lot of time for chart checking, which leads to a delay of restoration. To achieve a rapid restoration, the abnormal part assignment system (APAS), which can specify abnormality instantaneously even by a common operator, was developed. It has been confirmed that with this system, rapid initial response and prompt restoration can be effectively made. (A.O.)

  10. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-04-01

    In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure and a simultaneous loss of forced cool of the core

  11. Features of FJR 710 engine

    Energy Technology Data Exchange (ETDEWEB)

    Miyatake, Hirokazu; Kobayashi, Hideo [Ishikawajima-Harima Heavy Industries Co., Ltd., Tokyo (Japan)

    1989-03-01

    The FJR 710/600 high by-pass ratio turbofan engine was outlined, which was developed under the National Research and Development Program of Agency of Industrial Science and Technology, MITI dated from 1971. Results of developmental tests and flying tests of the engine were reported, which were performed to install the engine on the quiet STOL research aircraft under development by the National Aerospace Laboratory (NAL), the Science and Technology Agency, and the STOL aircraft was also outlined. Developmental tests of FJR 710 (design specification : thrust of 50kN, specific fuel consumption of 37.4g/N.h, turbine inlet temperature of 1.250{degree}C) were performed for 4,800 hours to verify airworthiness, and from those test results, the results of such environmental resistance tests as a foreign object strike test, icing test, inlet distortion test, cross wind test, and noise and exhaust measurement were reported in detail. 20 flying tests by a flying test bed (FTB) aircraft and flight tests by the STOL aircraft were also outlined. 10 refs., 5 figs., 2 tabs.

  12. 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.

  13. Integral fast reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  14. Integral Fast Reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  15. Main engineering features driving design concept and engineering design constraints

    International Nuclear Information System (INIS)

    Saito, Ryusei; Kobayashi, Takeshi; Yamada, Masao

    1987-09-01

    Major engineering design philosophies are described, which are essential bases for an engineering design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, engineering design drivers and engineering design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as coil system, a mechanical configuration, a tritium breeding scenario, etc.. The design constraints may follow a natural law or engineering limit, such as material strength, coil current density, and so on. (author)

  16. Safety engineering experiments of explosives

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Noboru

    1987-07-24

    The outline of large scale experiments carried out every year since 1969 to obtain fundamental data and then establish the safety engineering standards concerning the manufacturing, storage and transportation, etc. of all explosives was described. Because it becomes recently difficult to ensure the safety distance in powder magazines and powder plants, the sandwich structure with sand is thought to be suitable as the neighboring barrier walls. The special vertical structure for embankments to provide against a emergency explosion is effective to absorb the blast. Explosion behaviors such as initiating sensitivity, detonation, sympathetic detonation, and shock occurence of the ANFO explosives in place of dynamite and the slurry explosives were studied. The safety engineering standards for the manufacturing and application of explosives were studied to establish because accidents by tabacco fire are not still distinguished. Much data concerning early stage fire fighting, a large quantity of flooding and shock occurence from a assumption of ignition during machining in the propellants manufacturing plant, could be obtained. Basic studies were made to prevent pollution in blasting sites. Collected data are utilized for the safety administration after sufficient discussion. (4 figs, 2 tabs, 3 photos, 17 refs)

  17. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1995-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  18. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1994-01-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

  19. Safety features of subcritical fluid fueled systems

    Energy Technology Data Exchange (ETDEWEB)

    Bell, C.R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

  20. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-09-01

    The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

  1. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-01-01

    This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

  2. Engineering features of the INTOR conceptual design

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1981-01-01

    The INTOR engineering design has been strongly influenced by considerations for assembly and maintenance. A maintenance philosophy was established at the outset of the conceptual design to insure that the tokamak configuration would be developed to accommodate maintenance requirements. The main features of the INTOR design are summarized in this paper with primary emphasis on the impact of maintenance considerations

  3. Space transportation main engine reliability and safety

    Science.gov (United States)

    Monk, Jan C.

    1991-01-01

    Viewgraphs are used to illustrate the reliability engineering and aerospace safety of the Space Transportation Main Engine (STME). A technology developed is called Total Quality Management (TQM). The goal is to develop a robust design. Reducing process variability produces a product with improved reliability and safety. Some engine system design characteristics are identified which improves reliability.

  4. X-ray and nuclear radiation facilities: personnel safety features

    International Nuclear Information System (INIS)

    Mason, W.J.; Pipes, E.W.; Rucker, T.R.; Smith, D.N.; West, C.M.

    1976-10-01

    The Oak Ridge Y-12 Plant is a research and production installation. The nature and versatility of this work require the use of a large number and variety of x-ray and radiographic sources for nondestructive testing and material analyses. Presently, there are over 80 x-ray generators in the plant, which range in size from small, portable units which operate at a less than 50 kilovolts potential and 0.1 milliampere current to an electron linear accelerator which operates at 12-million electron volts and produces a radiation beam of such intensity that it could deliver a lethal dose to man in a fraction of a minute. There are also almost 50 gamma and neutron sources in use in the plant. These units range in size from a few millicuries to several hundred curies. Although the radiation safety at each of these facilities was considered adequate, the administrative and maintenance procedures became unduly complicated. Accordingly, engineering standards and uniform operating procedures were considered necessary to alleviate these complications and, in so doing, provide an improved measure of radiation safety. Development and implementation of these standards are described and the general philosophy and approach to these standards are outlined. Use of a matrix (type of installation versus radiation safety feature) to facilitate equipment classification and personnel safety feature requirements is presented. Included is a set of the standards showing formats, matrices, etc., and the detailed standards for each safety feature

  5. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-01-01

    The Modular High Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry, and the utilities. Near-term development is focused on electricity generation. The top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. Quantitatively, this requires that the design meet the US Environmental Protection Agency's Protective Action Guides at the site boundary and hence preclude the need for sheltering or evacuation of the public. To meet these stringent safety requirements and at the same time provide a cost competitive design requires the innovative use of the basic high temperature gas-cooled reactor features of ceramic fuel, helium coolant, and a graphite moderator. The specific fuel composition and core size and configuration have been selected to the use the natural characteristics of these materials to develop significantly higher margins of safety. In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure of a simultaneous loss of forced cooling of the core. The challenges to control of heat generation consider not only the failure to insert the reactivity control systems but also the withdrawal of control rods. Finally, challenges to control of chemical attack of the ceramic-coated fuel are considered, including catastrophic failure of the steam generator, which allows water ingress, or failure of the pressure vessels, which allows air ingress. The plant's response to these extreme challenges is not dependent on operator action, and the events considered encompass conceivable operator errors

  6. Safety design features of the IRIS

    International Nuclear Information System (INIS)

    2009-01-01

    The International Reactor Innovative and Secure (IRIS) is an advanced, integral, light water cooled reactor of medium generating capacity (335 MW(e)), that features an integral reactor vessel containing all the reactor primary system components, including steam generators, coolant pumps, pressurizer and heaters, and control rod drive mechanisms; in addition to the typical core, internals, control rods and neutron reflector. This integral configuration allows for the use of a small, high design pressure, spherical steel containment which results in a significant reduction in the size of the nuclear island. Other IRIS innovations include a simplified passive safety system concept and equipment features that derive from the 'safety-by-design' philosophy. This design approach allows for elimination of certain accident initiators at the design stage, or when outright elimination is not possible, decreases accident consequences and/or their probability of occurrence. Major design characteristics of the IRIS are given. As part of the IRIS pre-application licensing review by the U.S. Nuclear Regulatory Commission (NRC), the IRIS design team has developed a test plan that will provide the necessary data for safety analysis computer model verification, as well as for verifying the manufacturing feasibility, operability, and durability of new component designs

  7. Monitor for safety engineering facility

    International Nuclear Information System (INIS)

    Sato, Akira; Kaneda, Mitsunori.

    1982-01-01

    Purpose: To improve the reactor safety and decrease misoperation upon periodical inspection by instantly obtaining the judgement for the stand-by states in engineering safety facilities of a nuclear power plant. Constitution: Process inputs representing the states of valves, pumps, flowrates or the likes of the facility are gathered into an input device and inputted to a status monitor. The status of the facility inputted to the input device are judged for each of the inputs in a judging section and recognized as a present system stand-by pattern of the system (Valve) to be inspected. While on the other hand, a normal system stand-by pattern previously stored in a memory unit is read out by an instruction from an operator console and judged by comparison with the system stand-by pattern in a comparison section. The results are displayed on a display device. Upon periodical inspection, inspection procedures stored in the memory unit are displayed on the display device by the instruction from the operator console. (Seki, T.)

  8. Modeling crash injury severity by road feature to improve safety.

    Science.gov (United States)

    Penmetsa, Praveena; Pulugurtha, Srinivas S

    2018-01-02

    The objective of this research is 2-fold: to (a) model and identify critical road features (or locations) based on crash injury severity and compare it with crash frequency and (b) model and identify drivers who are more likely to contribute to crashes by road feature. Crash data from 2011 to 2013 were obtained from the Highway Safety Information System (HSIS) for the state of North Carolina. Twenty-three different road features were considered, analyzed, and compared with each other as well as no road feature. A multinomial logit (MNL) model was developed and odds ratios were estimated to investigate the effect of road features on crash injury severity. Among the many road features, underpass, end or beginning of a divided highway, and on-ramp terminal on crossroad are the top 3 critical road features. Intersection crashes are frequent but are not highly likely to result in severe injuries compared to critical road features. Roundabouts are least likely to result in both severe and moderate injuries. Female drivers are more likely to be involved in crashes at intersections (4-way and T) compared to male drivers. Adult drivers are more likely to be involved in crashes at underpasses. Older drivers are 1.6 times more likely to be involved in a crash at the end or beginning of a divided highway. The findings from this research help to identify critical road features that need to be given priority. As an example, additional advanced warning signs and providing enlarged or highly retroreflective signs that grab the attention of older drivers may help in making locations such as end or beginning of a divided highway much safer. Educating drivers about the necessary skill sets required at critical road features in addition to engineering solutions may further help them adopt safe driving behaviors on the road.

  9. Demonstration of inherent safety features of HTGRs using the HTTR

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

    2004-01-01

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

  10. OASIS: An automotive analysis and safety engineering instrument

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  11. Maintenance of civil engineering structures important to safety of Nuclear Power Plants

    International Nuclear Information System (INIS)

    2002-03-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. This safety standard is written to specify the objectives and minimum requirements for the design of civil engineering buildings/structures that are to be fulfilled to provide adequate assurance for safety of nuclear installations in India

  12. Tools for plant safety engineer

    International Nuclear Information System (INIS)

    Fabic, S.

    1996-01-01

    This paper contains: - review of tools for monitoring plant safety equipment reliability and readiness, before and accident (performance indicators for monitoring the risk and reliability performance and for determining when degraded performance alert levels are achieved) - brief reviews of tools for use during an accident: Emergency Operating Procedures (EOPs), Emergency Response Data System (ERDS), Reactor Safety Assessment System (RSAS), Computerized Accident Management Support

  13. Inherent safety features in balance-of-plant layout

    International Nuclear Information System (INIS)

    Wattelet, P.L.; Green, K.J.

    1992-01-01

    Future nuclear units must be more economical to construct and operate, and, at the same time, clearly incorporate advances in safety over the current generation of light water reactors. To achieve these goals, the root causes of safety issues must be addressed. In this way, global, cost-effective solutions can be implemented. With simple, direct design approaches, the licensing risk is minimized and configuration control is enhanced. With proper planning in the early stages of plant design, postulated accidents and events can often be mitigated by passive features inherent in the basic structure and layout, eliminating expensive added protective structures and components often found in current designs. Korea Electric Power Corporation's Yonggwang (YGN) Units 3 and 4, shown in an artist's rendering in Figure 1, are now under construction in Korea. Engineering is more than 85% complete, and Unit 3 construction is more than 50% complete. Significant steps toward design simplification and safety enhancement have been made by addressing safety concerns very early in the design effort. The tools used to achieve this were improved symmetry and separation, isolation of potential hazards, and an improved design process

  14. Engineering design guidelines for nuclear criticality safety

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1988-08-01

    This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

  15. ENGINEERED BARRIER SYSTEM FEATURES, EVENTS AND PROCESSES

    Energy Technology Data Exchange (ETDEWEB)

    Jaros, W.

    2005-08-30

    The purpose of this report is to evaluate and document the inclusion or exclusion of engineered barrier system (EBS) features, events, and processes (FEPs) with respect to models and analyses used to support the total system performance assessment for the license application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical basis for exclusion screening decisions. This information is required by the U.S. Nuclear Regulatory Commission (NRC) at 10 CFR 63.114 (d, e, and f) [DIRS 173273]. The FEPs addressed in this report deal with those features, events, and processes relevant to the EBS focusing mainly on those components and conditions exterior to the waste package and within the rock mass surrounding emplacement drifts. The components of the EBS are the drip shield, waste package, waste form, cladding, emplacement pallet, emplacement drift excavated opening (also referred to as drift opening in this report), and invert. FEPs specific to the waste package, cladding, and drip shield are addressed in separate FEP reports: for example, ''Screening of Features, Events, and Processes in Drip Shield and Waste Package Degradation'' (BSC 2005 [DIRS 174995]), ''Clad Degradation--FEPs Screening Arguments (BSC 2004 [DIRS 170019]), and Waste-Form Features, Events, and Processes'' (BSC 2004 [DIRS 170020]). For included FEPs, this report summarizes the implementation of the FEP in the TSPA-LA (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical basis for exclusion from TSPA-LA (i.e., why the FEP is excluded). This report also documents changes to the EBS FEPs list that have occurred since the previous versions of this report. These changes have resulted due to a reevaluation of the FEPs for TSPA-LA as identified in Section 1.2 of this report and described in more detail in Section 6.1.1. This revision addresses updates in Yucca Mountain Project

  16. ENGINEERED BARRIER SYSTEM FEATURES, EVENTS AND PROCESSES

    International Nuclear Information System (INIS)

    Jaros, W.

    2005-01-01

    The purpose of this report is to evaluate and document the inclusion or exclusion of engineered barrier system (EBS) features, events, and processes (FEPs) with respect to models and analyses used to support the total system performance assessment for the license application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical basis for exclusion screening decisions. This information is required by the U.S. Nuclear Regulatory Commission (NRC) at 10 CFR 63.114 (d, e, and f) [DIRS 173273]. The FEPs addressed in this report deal with those features, events, and processes relevant to the EBS focusing mainly on those components and conditions exterior to the waste package and within the rock mass surrounding emplacement drifts. The components of the EBS are the drip shield, waste package, waste form, cladding, emplacement pallet, emplacement drift excavated opening (also referred to as drift opening in this report), and invert. FEPs specific to the waste package, cladding, and drip shield are addressed in separate FEP reports: for example, ''Screening of Features, Events, and Processes in Drip Shield and Waste Package Degradation'' (BSC 2005 [DIRS 174995]), ''Clad Degradation--FEPs Screening Arguments (BSC 2004 [DIRS 170019]), and Waste-Form Features, Events, and Processes'' (BSC 2004 [DIRS 170020]). For included FEPs, this report summarizes the implementation of the FEP in the TSPA-LA (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical basis for exclusion from TSPA-LA (i.e., why the FEP is excluded). This report also documents changes to the EBS FEPs list that have occurred since the previous versions of this report. These changes have resulted due to a reevaluation of the FEPs for TSPA-LA as identified in Section 1.2 of this report and described in more detail in Section 6.1.1. This revision addresses updates in Yucca Mountain Project (YMP) administrative procedures as they

  17. Detonation Jet Engine. Part 2--Construction Features

    Science.gov (United States)

    Bulat, Pavel V.; Volkov, Konstantin N.

    2016-01-01

    We present the most relevant works on jet engine design that utilize thermodynamic cycle of detonative combustion. Detonation engines of various concepts, pulse detonation, rotational and engine with stationary detonation wave, are reviewed. Main trends in detonation engine development are discussed. The most important works that carried out…

  18. Education and Training of Safety Regulation for Nuclear Safety Infrastructure: Its Necessity and Unique Features

    International Nuclear Information System (INIS)

    Choi, Young Sung; Choi, Young Joon; Lee, Jae Cheon

    2009-01-01

    Faced with global warming and electricity demands, countries over the world recognize the comparative advantages of nuclear energy. It is estimated that about 300 nuclear power plants (NPPs) expect to be constructed until 2030 worldwide. In addition, according to the IAEA, approximately 20 new countries might have their first NPP in operation by 2030 in the high projection compared with bout 5 new countries in the low projection. When introducing nuclear power, the implementation of an appropriate infrastructure to address all of the relevant issues is a central concern of international community. In particular, nuclear power program requires, at an earlier stage than when construction starts, the development of a legal and regulatory framework and training of regulators and safety experts whose combined knowledge adequately covers all areas of nuclear safety and regulation applied at a NPP construction and operation. As an essential component of such human resource development, special attention was paid to the provision of education and training to regulators of which countries plan to introduce NPPs. In term of education theory, safety regulation has some unique features in learning and teaching, which are different from those of nuclear engineering or development. This paper overviews nuclear safety infrastructure, explores the roles of exporting countries, and presents features and components in education of nuclear safety regulation

  19. Modeling for safety in a synthesis-centric systems engineering framework

    NARCIS (Netherlands)

    Markovski, J.; Mortel - Fronczak, van de J.M.; Ortmeier, F.; Daniel, P.

    2012-01-01

    The ever-increasing complexity of safety-critical systems puts high demands on safety assurance and certification. We focus on the development of control software, where safety) requirements engineering plays a crucial and delicate role. Nowadays, most of the safety features are ensured by the

  20. 46. The goals of safety engineering department of the plant

    International Nuclear Information System (INIS)

    Ivanov, A.V.

    1993-01-01

    The goals of safety engineering department of the plant, including elaboration of instructions on safety engineering on all specialities, safety engineering training of all labours working on the plant and control for abidance by the instructions on safety engineering were discussed.

  1. 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

  2. Criticality safety engineer training at WSRC

    International Nuclear Information System (INIS)

    Williamson, T.G.; Mincey, J.F.

    1993-01-01

    Two programs designed to prepare engineers for certification as criticality safety engineers are offered at Westinghouse Savannah River Company (WSRC). One program, Student On Loan Criticality Engineer Training (SOLCET), is an intensive 2-yr course involving lectures, rigorous problem assignments, and mentoring. The other program, In-Field Criticality Engineer Training (IN-FIELD), is a less intensive series of lectures and problem assignments. Both courses are conducted by members of the Applied Physics Group (APG) of the Savannah River Technical Center, the organization at WSRC responsible for the operation and maintenance of criticality codes and for training of code users

  3. Developing safety culture in nuclear power engineering

    International Nuclear Information System (INIS)

    Tevlin, S.A.

    2000-01-01

    The new issue (no. 11) of the IAEA publications series Safety Reports, devoted to the safety culture in nuclear engineering Safety culture development in the nuclear activities. Practical recommendations to achieve success, is analyzed. A number of recommendations of international experts is presented and basic general indicators of satisfactory and insufficient safety culture in the nuclear engineering are indicated. It is shown that the safety culture has two foundations: human behavior and high quality of the control system. The necessity of creating the confidence by the management at all levels of the enterprise, development of individual initiative and responsibility of the workers, which make it possible to realize the structural hierarchic system, including technical, human and organizational constituents, is noted. Three stages are traced in the process of introducing the safety culture. At the first stage the require,emts of scientific-technical documentation and provisions of the governmental, regional and control organs are fulfilled. At the second stage the management of the organization accepts the safety as an important direction in its activities. At the third stage the organization accomplishes its work, proceeding from the position of constant safety improvement. The general model of the safety culture development is considered [ru

  4. Engineering safety assessment. An introduction

    International Nuclear Information System (INIS)

    Thomson, J.R.

    1987-01-01

    There have been several major industrial accidents in recent years, notably the Chernobyl reactor accident, the Bhopal disaster and the Mexico City gas explosions. It is essential that engineers should understand the natures of the hazards and risks of industrial plant. This book is intended to give an introduction to this subject. Determining the risks associated with hazardous plant such as a chemical plant or a nuclear power station requires knowledge of the probability of a given accident, as well as knowledge of the likely consequences in terms of human lives or damaged property. In this book, the author explains these two main themes in assessing overall risk. He introduces basic probability theory, the assessment of systems reliabilities and structural reliabilities, and the means used to estimate the consequences of hypothetical major accidents. In particular, the methodologies for estimating the consequences of accidental explosion and accidents involving the release of airborne toxic materials (including chemical toxins, carcinogens, and radioactive materials) are explained. A final chapter explores the difficult question of 'acceptable' risk, and how much can justifiably be spent on risk reduction. Intended for use by final year students of chemical, mechanical or nuclear engineering, this book will also be suitable for postgraduate students and practising engineers and scientists whose work involves them in risk assessment. (author)

  5. Patient safety trilogy: perspectives from clinical engineering.

    Science.gov (United States)

    Gieras, Izabella; Sherman, Paul; Minsent, Dennis

    2013-01-01

    This article examines the role a clinical engineering or healthcare technology management (HTM) department can play in promoting patient safety from three different perspectives: a community hospital, a national government health system, and an academic medical center. After a general overview, Izabella Gieras from Huntington Hospital in Pasadena, CA, leads off by examining the growing role of human factors in healthcare technology, and describing how her facility uses clinical simulations in medical equipment evaluations. A section by Paul Sherman follows, examining patient safety initiatives from the perspective of the Veterans Health Administration with a focus on hazard alerts and recalls. Dennis Minsent from Oregon Health & Science University writes about patient safety from an academic healthcare perspective, and details how clinical engineers can engage in multidisciplinary safety opportunities.

  6. Adding a visualization feature to web search engines: it's time.

    Science.gov (United States)

    Wong, Pak Chung

    2008-01-01

    It's widely recognized that all Web search engines today are almost identical in presentation layout and behavior. In fact, the same presentation approach has been applied to depicting search engine results pages (SERPs) since the first Web search engine launched in 1993. In this Visualization Viewpoints article, I propose to add a visualization feature to Web search engines and suggest that the new addition can improve search engines' performance and capabilities, which in turn lead to better Web search technology.

  7. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

  8. Improving Safety through Human Factors Engineering.

    Science.gov (United States)

    Siewert, Bettina; Hochman, Mary G

    2015-10-01

    Human factors engineering (HFE) focuses on the design and analysis of interactive systems that involve people, technical equipment, and work environment. HFE is informed by knowledge of human characteristics. It complements existing patient safety efforts by specifically taking into consideration that, as humans, frontline staff will inevitably make mistakes. Therefore, the systems with which they interact should be designed for the anticipation and mitigation of human errors. The goal of HFE is to optimize the interaction of humans with their work environment and technical equipment to maximize safety and efficiency. Special safeguards include usability testing, standardization of processes, and use of checklists and forcing functions. However, the effectiveness of the safety program and resiliency of the organization depend on timely reporting of all safety events independent of patient harm, including perceived potential risks, bad outcomes that occur even when proper protocols have been followed, and episodes of "improvisation" when formal guidelines are found not to exist. Therefore, an institution must adopt a robust culture of safety, where the focus is shifted from blaming individuals for errors to preventing future errors, and where barriers to speaking up-including barriers introduced by steep authority gradients-are minimized. This requires creation of formal guidelines to address safety concerns, establishment of unified teams with open communication and shared responsibility for patient safety, and education of managers and senior physicians to perceive the reporting of safety concerns as a benefit rather than a threat. © RSNA, 2015.

  9. Structural safety features for superconducting magnets

    International Nuclear Information System (INIS)

    Lehner, J.; Reich, M.; Powell, J.; Bezler, P.; Gardner, D.; Yu, W.; Chang, T.Y.

    1975-01-01

    A survey has been carried out for various potential structural safety problems of superconducting fusion magnets. These areas include: (1) Stresses due to inhomogeneous temperature distributions in magnets where normal regions have been initiated. (2) Stress distributions and yield forces due to cracks and failed regions. (3) Superconducting magnet response due to seismic excitation. These analyses have been carried out using a variety of large capacity finite element computer codes that allow for the evaluation of stresses in elastic or elastic-plastic zones and around singularities in the magnet structure. Thus far, these analyses have been carried out on UWMAK-I type magnet systems

  10. 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

  11. The safety features of an integrated maritime reactor

    International Nuclear Information System (INIS)

    Miyakoshi, Junichi; Yamada, Nobuyuki; Kuwahara, Shin-ichi

    1975-01-01

    The EFDR-80, a typical integrated maritime reactor, which is being developed in West Germany is outlined. The safety features of the integrated maritime reactor are presented with the analysis of reactor accidents and hazards, and are compared with those of the separated maritime reactor. Furthermore, the safety criteria of maritime reactors in Japan and West Germany are compared, and some of the differences are presented from the viewpoint of reactor design and safety analysis. In this report the authors express an earnest desire that the definite and reasonable safety criteria of the integrated maritime reactor should be established and that the safety criteria of the nuclear ship should be standardized internationally. (auth.)

  12. Integrated system of safety features for spent fuel interim storage

    International Nuclear Information System (INIS)

    Pantazi, Doina; Stanciu, Marcela; Mateescu, Silvia; Marin, Ion

    1999-01-01

    The design of the spent fuel interim storage facility (SFISF) must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility. To elaborate the safety documentation necessary for licensing, we were trying to chose the most appropriate approach related to safety features for SFISF, based on national and international regulations, standards and recommendations, as well as on the experience of other countries with similar facilities and finally, on our own experience in designing other nuclear objectives in Romania. The paper presents the issues that we consider important for the safety evaluation and are developed as a detailed diagram. The diagram contains in a logical succession the following issues: - fundamental principles of radioprotection; - fundamental safety principles of radioactive waste management; - safety objectives of SFISF; - safety criteria for SFISF; - safety requirements for SFISF; - siting criteria for SFISF; - siting requirements for SFISF. (authors)

  13. Adequacy features of Nucleoelectrica Argentina Safety Management

    Energy Technology Data Exchange (ETDEWEB)

    Rapoport, H [Nucleoelectrica Argentina S.A., Buenso Aires (Argentina)

    1997-12-01

    The Argentine Nuclear Power Plants (NPP) ATUCHA I (357 Mw, SIEMENS) and EMBALSE (648 Mw, CANDU), both of PHWR type, were owned and operated until August 94 by the Comision Nacional de Energia Atomica (CNEA). Until that date, CNEA, the national agency for nuclear R and D, concentrated three activities or roles: Research, Nuclear Regulations and NPP Operation. Since August 1994 NPP`s are owned and operated by a state electrical company (Nucleoelectrica Argentina S.A.) the nuclear utility supplying 15% of the national electrical generation demand. NASA is going to be privatized according to a recent national law regulating nuclear activities. The transition from a research agency to a commercial company requires the introduction of changes in the management of activities. Mention of these changes is limited to those relating to Safety.

  14. Adequacy features of Nucleoelectrica Argentina Safety Management

    International Nuclear Information System (INIS)

    Rapoport, H.

    1997-01-01

    The Argentine Nuclear Power Plants (NPP) ATUCHA I (357 Mw, SIEMENS) and EMBALSE (648 Mw, CANDU), both of PHWR type, were owned and operated until August 94 by the Comision Nacional de Energia Atomica (CNEA). Until that date, CNEA, the national agency for nuclear R and D, concentrated three activities or roles: Research, Nuclear Regulations and NPP Operation. Since August 1994 NPP's are owned and operated by a state electrical company (Nucleoelectrica Argentina S.A.) the nuclear utility supplying 15% of the national electrical generation demand. NASA is going to be privatized according to a recent national law regulating nuclear activities. The transition from a research agency to a commercial company requires the introduction of changes in the management of activities. Mention of these changes is limited to those relating to Safety

  15. LFR safety features through intrinsic negative reactivity feedbacks

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2012-01-01

    The safety of Lead-cooled Fast Reactors can rely on intrinsic features such as: • the impossibility of Lead boiling, hence the unreliability of core (only) voiding; • the buoyancy of Control Rods in Lead, allowing their safe positioning also below the active region. For heightening the safety features of LFRs in safety analyses it could be required to approach the evaluation of the reactivity coefficients from a more physical point of view, including more elementary mechanisms, each one related to the proper driving temperature

  16. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  17. An engineer-constructor's view of nuclear power plant safety

    International Nuclear Information System (INIS)

    Landis, J.W.; Jacobs, S.B.

    1984-01-01

    At SWEC we have been involved in the development of safety features of nuclear power plants ever since we served as the engineer-constructur for the first commerical nuclear power station at Shippingport, Pennsylvania, in the 1950s. Our personnel have pioneered a number of safety innovations and improvements. Among these innovations is the subatmospheric containment for pressurized water reactor (PWR) power plants. This type of containment is designed so that leakage will terminate within 1 to 2 hours of the worst postulated loss of coolant accident. Other notable contributions include first use of reinforced-concrete atmospheric containments for PWR power plants and of reinforced-concrete, vapor-suppression containments for boiling water reactor (BWR) power plants. Both concepts meet rigorous U.S. safety requirements. SWEC has performed a substantial amount of work on developing standardized plant designs and has developed standardized engineering and construction techniques and procedures. Standardization concepts are being developed in Canada, France, USSR, and Germany, as well as in the United States. The West German convoy concept, which involves developing a number of standardized plants in a common effort, has been quite successful. We believe standardization contributes to safety in a number of ways. Use of standardized designs, procedures, techniques, equipment, and methods increases efficiency and results in higher quality. Standardization also reduces the design variations with which plant operators, emergency teams, and regulatory personnel must be familiar, thus increasing operator capability, and permits specialized talents to be focused on important safety considerations. (orig./RW)

  18. Reactor engineering and engineered reactor safety in France

    International Nuclear Information System (INIS)

    1987-01-01

    The proceedings give the full text of the lectures held by acknowledged French experts at the KTG Seminar in Mainz on March 10, 1987, all dealing with the leading topic of the current status of reactor engineering and development in France. Although the basic engineering principles and construction lines as well as the safety philosophy are the same in France as in West Germany, there have been distinctive developments over many years in the two countries that by now are not well known even among experts in this field, and hence cannot be properly assessed. Non-availability of relevant surveys or other type of literature in the German language reviewing the French developments is another factor that hitherto was a handicap to mutual exchange of information. The seminar was intended to close this gap. The proceedings should be read by all those in West Germany who wish to be informed about the developments in reactor engineering and reactor safety in France. (orig./DG) [de

  19. Atomic power engineering under falsified safety standards

    International Nuclear Information System (INIS)

    Ackerman, A.J.

    1974-01-01

    In July 1970 the United States Department of Justice accused the American Society of Mechanical Engineers (ASME) of violating the Sherman Antitrust Act and of acting in restraint of trade by restricting the ASME Certificate of Authorization and the use of the Code Symbol Stamps to boilers and pressure vessels manufactured in the United States and Canada. During the succeeding two years attorneys for the parties in the case formulated a Consent Decree without a public confrontation in the Court. Furthermore, the membership of ASME was kept uninformed until October of 1972, after the Consent and Final Judgment had become effective and new procedures had been developed for allowing foreign manufacturers to apply the ASME Code Symbol Stamps to their products. As a consequence, a breakdown in engineered safety standards has been sanctioned and this is undermining the engineering profession's overriding reponsibility to protect the public health and safety. This breakdown of professional responsibility is especially serious in the new technology of atomic power. American insurance companies, which have traditionally written 100% insurance coverage for property damage and third party liability against explosions of high pressure steam boilers bearing the ASME Code Stamp, have refused to write such insurance coverage on nuclear reactors. In the author's opinion there is evidence that the Consent was formulated under collusive proceedings and he calls on the members and the Council of ASME to appeal for dismissal of the Consent Decree. 24 refs

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  1. ENGINEERED BARRIER SYSTEM FEATURES, EVENTS, AND PROCESSES

    International Nuclear Information System (INIS)

    2005-01-01

    This analysis report is one of the technical reports containing documentation of the Environmental Radiation Model for Yucca Mountain, Nevada (ERMYN), a biosphere model supporting the total system performance assessment (TSPA) for the license application (LA) for the Yucca Mountain repository. This analysis report describes the development of biosphere dose conversion factors (BDCFs) for the volcanic ash exposure scenario, and the development of dose factors for calculating inhalation dose during volcanic eruption. A graphical representation of the documentation hierarchy for the ERMYN is presented in Figure 1 - 1. This figure shows the interrelationships among the products (i.e., analysis and model reports) developed for biosphere modeling and provides an understanding of how this analysis report contributes to biosphere modeling. This report is one of two reports that develop biosphere BDCFs, which are input parameters for the TSPA model. The ''Biosphere Model Report'' (BSC 2004 [DIRS 169460]) describes in detail the ERMYN conceptual model and mathematical model. The input parameter reports, shown to the right of the Biosphere Model Report in Figure 1-1, contain detailed descriptions of the model input parameters, their development and the relationship between the parameters and specific features, events and processes (FEPs). This report describes biosphere model calculations and their output, the BDCFs, for the volcanic ash exposure scenario. This analysis receives direct input from the outputs of the ''Biosphere Model Report'' (BSC 2004 [DIRS 169460]) and from the five analyses that develop parameter values for the biosphere model (BSC 2005 [DIRS 172827]; BSC 2004 [DIRS 169672]; BSC 2004 [DIRS 169673]; BSC 2004 [DIRS 169458]; and BSC 2004 [DIRS 169459]). The results of this report are further analyzed in the ''Biosphere Dose Conversion Factor Importance and Sensitivity Analysis'' (Figure 1 - 1). The objective of this analysis was to develop the BDCFs for the

  2. Fuel elements and safety engineering goals

    International Nuclear Information System (INIS)

    Schulten, R.; Bonnenberg, H.

    1990-01-01

    There are good prospects for silicon carbide anti-corrosion coatings on fuel elements to be realised, which opens up the chance to reduce the safety engineering requirements to the suitable design and safe performance of the ceramic fuel element. Another possibility offered is combined-cycle operation with high efficiencies, and thus good economic prospects, as with this design concept combining gas and steam turbines, air ingress due to turbine malfunction is an incident that can be managed by the system. This development will allow economically efficient operation also of nuclear power reactors with relatively small output, and hence contribute to reducing CO 2 emissions. (orig./DG) [de

  3. Physical principle and engineering features of the deep pool reactor for residential heating

    International Nuclear Information System (INIS)

    Shi Gong; Zhao Zhaoyi; Guo Jingren; Tian Jiafu

    1999-01-01

    The use of nuclear energy for low temperature heating is confronted with challenges of safety and economy. The deep pool reactor, a low temperature heating reactor based on novel design principles, has been studied in detail. Results show that it has excellent safety and economic features, and is very suitable for low temperature heating purposes. The whole heating system including the nuclear reactor will be a simple and easy engineering system with the characteristics of reliability, safety and economy because the system and all its devices are based on low temperature and ordinary pressure

  4. ENGINEERED BARRIER SYSTEM FEATURES, EVENTS, AND PROCESSES

    Energy Technology Data Exchange (ETDEWEB)

    na

    2005-05-30

    This analysis report is one of the technical reports containing documentation of the Environmental Radiation Model for Yucca Mountain, Nevada (ERMYN), a biosphere model supporting the total system performance assessment (TSPA) for the license application (LA) for the Yucca Mountain repository. This analysis report describes the development of biosphere dose conversion factors (BDCFs) for the volcanic ash exposure scenario, and the development of dose factors for calculating inhalation dose during volcanic eruption. A graphical representation of the documentation hierarchy for the ERMYN is presented in Figure 1 - 1. This figure shows the interrelationships among the products (i.e., analysis and model reports) developed for biosphere modeling and provides an understanding of how this analysis report contributes to biosphere modeling. This report is one of two reports that develop biosphere BDCFs, which are input parameters for the TSPA model. The ''Biosphere Model Report'' (BSC 2004 [DIRS 169460]) describes in detail the ERMYN conceptual model and mathematical model. The input parameter reports, shown to the right of the Biosphere Model Report in Figure 1-1, contain detailed descriptions of the model input parameters, their development and the relationship between the parameters and specific features, events and processes (FEPs). This report describes biosphere model calculations and their output, the BDCFs, for the volcanic ash exposure scenario. This analysis receives direct input from the outputs of the ''Biosphere Model Report'' (BSC 2004 [DIRS 169460]) and from the five analyses that develop parameter values for the biosphere model (BSC 2005 [DIRS 172827]; BSC 2004 [DIRS 169672]; BSC 2004 [DIRS 169673]; BSC 2004 [DIRS 169458]; and BSC 2004 [DIRS 169459]). The results of this report are further analyzed in the ''Biosphere Dose Conversion Factor Importance and Sensitivity Analysis'' (Figure 1 - 1). The

  5. Curriculum: Integrating Health and Safety Into Engineering Curricula.

    Science.gov (United States)

    Talty, John T.

    1985-01-01

    National Institute for Occupational Safety and Health instituted a project in 1980 to encourage engineering educators to focus on occupational safety and health issues in engineering curricula. Progress to date is outlined, considering specific results in curriculum development, engineering society interaction, and formation of a teaching…

  6. Integrating system safety into the basic systems engineering process

    Science.gov (United States)

    Griswold, J. W.

    1971-01-01

    The basic elements of a systems engineering process are given along with a detailed description of what the safety system requires from the systems engineering process. Also discussed is the safety that the system provides to other subfunctions of systems engineering.

  7. Design data and safety features of commerical nuclear power plant

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1976-06-01

    Design data, safety features, and site characteristics are summarized for 34 nuclear power units in 17 power stations in the United States. Six pages of data are presented for each plant, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. This volume covers Light Water Reactors (LWRs) with dockets 50-508 through 50-549, four HTGRs--50-171, 50-267, 50-450/451, 50-463/464, the Atlantic Floating Station 50-477/478, and the Clinch River Breeder 50-537

  8. Reactor protection system including engineered features actuation system

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

    The safety concept requires to ensure that - the reactor protection system - the active engineered safeguard - and the necessary auxiliary systems are so designed and interfaced in respect of design and mode of action that, in the event of single component failure reliable control of the consequences of accidents remains ensured at all times and that the availability of the power plant is not limited unnecessarily. In order to satisfy these requirements due, importance was attached to a consistent spacial separation of the mutually redundant subsystems of the active safety equipment. The design and layout of the reactor protection system, of the power supply (emergency power supply), and of the auxiliary systems important from the safety engineering point of view, are such that their subsystems also largely satisfy the requirements of independence and spacial separation. (orig./RW)

  9. Recent Development of the Two-Stroke Engine. II - Design Features. 2; Design Features

    Science.gov (United States)

    Zeman, J.

    1945-01-01

    Completing the first paper dealing with charging methods and arrangements, the present paper discusses the design forms of two-stroke engines. Features which largely influence piston running are: (a) The shape and surface condition of the sliding parts. (b) The cylinder and piston materials. (c) Heat conditions in the piston, and lubrication. There is little essential difference between four-stroke and two-stroke engines with ordinary pistons. In large engines, for example, are always found separately cast or welded frames in which the stresses are taken up by tie rods. Twin piston and timing piston engines often differ from this design. Examples can be found in many engines of German or foreign make. Their methods of operation will be dealt with in the third part of the present paper, which also includes the bibliography. The development of two-stroke engine design is, of course, mainly concerned with such features as are inherently difficult to master; that is, the piston barrel and the design of the gudgeon pin bearing. Designers of four-stroke engines now-a-days experience approximately the same difficulties, since heat stresses have increased to the point of influencing conditions in the piston barrel. Features which notably affect this are: (a) The material. (b) Prevailing heat conditions.

  10. Safety Features of Material and Personnel Movement Devices. Module SH-25. Safety and Health.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This student module on safety features of material and personnel movement devices is one of 50 modules concerned with job safety and health. This module covers safe conditions and operating practices for conveyors, elevators, escalators, moving walks, manlifts, forklifts, and motorized hand trucks. Following the introduction, 10 objectives (each…

  11. Experience with performance based training of nuclear criticality safety engineers

    International Nuclear Information System (INIS)

    Taylor, R.G.

    1993-01-01

    For non-reactor nuclear facilities, the U.S. Department of Energy (DOE) does not require that nuclear criticality safety engineers demonstrate qualification for their job. It is likely, however, that more formalism will be required in the future. Current DOE requirements for those positions which do have to demonstrate qualification indicate that qualification should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis is incompletely developed in some areas

  12. Safety risk management of underground engineering in China: Progress, challenges and strategies

    Directory of Open Access Journals (Sweden)

    Qihu Qian

    2016-08-01

    Full Text Available Underground construction in China is featured by large scale, high speed, long construction period, complex operation and frustrating situations regarding project safety. Various accidents have been reported from time to time, resulting in serious social impact and huge economic loss. This paper presents the main progress in the safety risk management of underground engineering in China over the last decade, i.e. (1 establishment of laws and regulations for safety risk management of underground engineering, (2 implementation of the safety risk management plan, (3 establishment of decision support system for risk management and early-warning based on information technology, and (4 strengthening the study on safety risk management, prediction and prevention. Based on the analysis of the typical accidents in China in the last decade, the new challenges in the safety risk management for underground engineering are identified as follows: (1 control of unsafe human behaviors; (2 technological innovation in safety risk management; and (3 design of safety risk management regulations. Finally, the strategies for safety risk management of underground engineering in China are proposed in six aspects, i.e. the safety risk management system and policy, law, administration, economy, education and technology.

  13. Investigation of road network features and safety performance.

    Science.gov (United States)

    Wang, Xuesong; Wu, Xingwei; Abdel-Aty, Mohamed; Tremont, Paul J

    2013-07-01

    The analysis of road network designs can provide useful information to transportation planners as they seek to improve the safety of road networks. The objectives of this study were to compare and define the effective road network indices and to analyze the relationship between road network structure and traffic safety at the level of the Traffic Analysis Zone (TAZ). One problem in comparing different road networks is establishing criteria that can be used to scale networks in terms of their structures. Based on data from Orange and Hillsborough Counties in Florida, road network structural properties within TAZs were scaled using 3 indices: Closeness Centrality, Betweenness Centrality, and Meshedness Coefficient. The Meshedness Coefficient performed best in capturing the structural features of the road network. Bayesian Conditional Autoregressive (CAR) models were developed to assess the safety of various network configurations as measured by total crashes, crashes on state roads, and crashes on local roads. The models' results showed that crash frequencies on local roads were closely related to factors within the TAZs (e.g., zonal network structure, TAZ population), while crash frequencies on state roads were closely related to the road and traffic features of state roads. For the safety effects of different networks, the Grid type was associated with the highest frequency of crashes, followed by the Mixed type, the Loops & Lollipops type, and the Sparse type. This study shows that it is possible to develop a quantitative scale for structural properties of a road network, and to use that scale to calculate the relationships between network structural properties and safety. Copyright © 2013 Elsevier Ltd. All rights reserved.

  14. Key features and progress of the KSTAR tokamak engineering

    International Nuclear Information System (INIS)

    Bak, J.S.; Choi, C.H.; Oh, Y.K.

    2003-01-01

    Substantial progress of the KSTAR tokamak engineering has been made on major tokamak structures, superconducting magnets, in-vessel components, diagnostic system, heating system, and power supplies. The engineering design has been elaborated to the extent necessary to allow a realistic assessment of its feasibility, performance, and cost. The prototype fabrication has been carried out to establish the reliable fabrication technologies and to confirm the validation of analyses employed for the KSTAR design. The completion of experimental building with beneficial occupancy for machine assembly was accomplished in Sep. 2002. The construction of special utility such as cryo-plant, de-ionized water-cooling system, and main power station will begin upon completion of building construction. The commissioning, construction, fabrication, and assembly of the whole facility will be going on by the end of 2005. This paper describes the main design features and engineering progress of the KSTAR tokamak, and elaborates the work currently underway. (author)

  15. 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.

  16. The design and safety features of the IRIS reactor

    International Nuclear Information System (INIS)

    Carelli, Mario D.; Conway, L.E.; Oriani, L.; Petrovic, B.; Lombardi, C.V.; Ricotti, M.E.; Barroso, A.C.O.; Collado, J.M.; Cinotti, L.; Todreas, N.E.; Grgic, D.; Moraes, M.M.; Boroughs, R.D.; Ninokata, H.; Ingersoll, D.T.; Oriolo, F.

    2004-01-01

    Salient features of the International Reactor Innovative and Secure (IRIS) are presented here. IRIS, an integral, modular, medium size (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. Described here are the features of the integral design which includes steam generators, pumps and pressurizer inside the vessel, together with the core, control rods, and neutron reflector/shield. A brief summary is provided of the IRIS approach to extended maintenance over a 48-month schedule. The unique IRIS safety-by-design approach is discussed, which, by eliminating accidents, at the design stage, or decreasing their consequences/probabilities when outright elimination is not possible, provides a very powerful first level of defense in depth. The safety-by-design allows a significant reduction and simplification of the passive safety systems, which are presented here, together with an assessment of the IRIS response to transients and postulated accidents

  17. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  18. 2012 national state safety engineers and traffic engineers peer-to-peer workshop.

    Science.gov (United States)

    2013-11-01

    The Illinois Department of Transportation (IDOT) and the Illinois Center for Transportation (ICT) sponsored and hosted the : 2012 National State Safety Engineers and Traffic Engineers Peer-to-Peer Workshop on November 14 and 15, 2012, at the : Hyatt ...

  19. Do Undergraduate Engineering Faculty Include Occupational and Public Health and Safety in the Engineering Curriculum?

    Science.gov (United States)

    Farwell, Dianna; And Others

    1995-01-01

    The purpose of this study was to determine whether and, if so, why engineering faculty include occupational and public health and safety in their undergraduate engineering courses. Data were collected from 157 undergraduate engineering faculty from 65 colleges of engineering in the United States. (LZ)

  20. Specific features of medicines safety and pharmacovigilance in Africa

    Science.gov (United States)

    Pal, Shanthi N.; Olsson, Sten; Dodoo, Alexander; Bencheikh, Rachida Soulayami

    2012-01-01

    The thalidomide tragedy in the late 1950s and early 1960s served as a wakeup call and raised questions about the safety of medicinal products. The developed countries rose to the challenge putting in place systems to ensure the safety of medicines. However, this was not the case for low-resource settings because of prevailing factors inherent in them. This paper reviews some of these features and the current status of pharmacovigilance in Africa. The health systems in most of the 54 countries of Africa are essentially weak, lacking in basic infrastructure, personnel, equipment and facilities. The recent mass deployment of medicines to address diseases of public health significance in Africa poses additional challenges to the health system with notable safety concerns. Other safety issues of note include substandard and counterfeit medicines, medication errors and quality of medicinal products. The first national pharmacovigilance centres established in Africa with membership of the World Health Organization (WHO) international drug monitoring programme were in Morocco and South Africa in 1992. Of the 104 full member countries in the programme, there are now 24 African countries with a further nine countries as associate members. The pharmacovigilance systems operational in African countries are based essentially on spontaneous reporting facilitated by the introduction of the new tool Vigiflow. The individual case safety reports committed to the WHO global database (Vigibase) attest to the growth of pharmacovigilance in Africa with the number of reports rising from 2695 in 2000 to over 25,000 in 2010. There is need to engage the various identified challenges of the weak pharmacovigilance systems in the African setting and to focus efforts on how to provide resources, infrastructure and expertise. Raising the level of awareness among healthcare providers, developing training curricula for healthcare professionals, provisions for paediatric and geriatric

  1. Safety features and research needs of westinghouse advanced reactors

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. Improved safety features in the design of Alto Lazio NPP

    International Nuclear Information System (INIS)

    Bava, G.; Cianciolo, T.; Del Nero, G.

    1988-01-01

    The ALTO LAZIO Nuclear Power Plant, two 1000Mwe units, is a BWR 6/MARK III located about 100 km north of Rome, on the Tyrrhenian Sea Coasts. The construction of the plant started in 1978, but it has recently been stopped by a Government decision following a national referendum, when the units were about 70% completed. This paper is mainly intended to illustrate the major safety features which have been implemented as result of specific requirements issued by the safety authority (ENEA DISP) during the construction permit stage or the subsequent licensing process. One of the tools used to identify the need for design modifications has been a comprehensive reliability analysis of safety system: in the paper the methods used and the major results obtained by this study are briefly presented. Also, the approach used in the investigation of severe accidents and major applications in the area of plant design and emergency procedures are briefly discussed; furthermore the trend toward a simpler mitigation concept is described

  3. Automated Flight Safety Inference Engine (AFSIE) System, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We propose to develop an innovative Autonomous Flight Safety Inference Engine (AFSIE) system to autonomously and reliably terminate the flight of an errant launch...

  4. Russia power engineering and power safety

    International Nuclear Information System (INIS)

    D'yakov, A.F.

    1995-01-01

    Results of work of the International consultative meeting: Russian-Europe: strategy of energy safety is described. The purpose of the meeting consisted in discussion of energy situation in Russia and Europe, prospects for provision of reliability, efficiency and safety of fuel and power supply in Russia and the role of the Russian fuel and power resonances in energy supply of Europe. The reporters at the meeting dealt with various aspects related to energy safety

  5. Engineering systems reliability, safety, and maintenance an integrated approach

    CERN Document Server

    Dhillon, B S

    2017-01-01

    Today, engineering systems are an important element of the world economy and each year billions of dollars are spent to develop, manufacture, operate, and maintain various types of engineering systems around the globe. Many of these systems are highly sophisticated and contain millions of parts. For example, a Boeing jumbo 747 is made up of approximately 4.5 million parts including fasteners. Needless to say, reliability, safety, and maintenance of systems such as this have become more important than ever before.  Global competition and other factors are forcing manufacturers to produce highly reliable, safe, and maintainable engineering products. Therefore, there is a definite need for the reliability, safety, and maintenance professionals to work closely during design and other phases. Engineering Systems Reliability, Safety, and Maintenance: An Integrated Approach eliminates the need to consult many different and diverse sources in the hunt for the information required to design better engineering syste...

  6. Psychological Safety and Norm Clarity in Software Engineering Teams

    OpenAIRE

    Lenberg, Per; Feldt, Robert

    2018-01-01

    In the software engineering industry today, companies primarily conduct their work in teams. To increase organizational productivity, it is thus crucial to know the factors that affect team effectiveness. Two team-related concepts that have gained prominence lately are psychological safety and team norms. Still, few studies exist that explore these in a software engineering context. Therefore, with the aim of extending the knowledge of these concepts, we examined if psychological safety and t...

  7. Potential safety features and safety analysis aspects for high performance light water reactor (HPLWR)

    International Nuclear Information System (INIS)

    Aksan, N.; Schulenberg, T.; Squarer, D.

    2003-01-01

    Research Activities are ongoing worldwide to develop advanced nuclear power plants with high thermal efficiency for the purpose to improve their economical competitiveness. Within the 5th Framework Programme of the European Commission, a project has been launched with the main objective to assess the technical and economical feasibility of a high efficiency LWR operating at super critical pressure conditions. Several European research institutions, industrial partners and the University of Tokyo participated and worked in this common research project. Within the aims of the development of the HPLWR is to use both passive and active safety systems for performing safety related functions in the event of transients or accidents. Consequently substantial effort has been invested in order to define the safety features of the plant in a European environment, as well as to incorporate passive safety features into the design. Throughout this process, the European Utility Requirements (EUR) and requirements known from Generation IV initiative were considered as a guideline in general terms in order to include further advanced ideas. The HPLWR general features were compared to both requirements, indicating a potential to meet these. Since, the supercritical HPLWR represents a challenge for best-estimate safety codes like RELAP5, CATHARE and TRAB due to the fact that these codes were developed for two-phase or single-phase coolant at pressures far below critical point, work on the preliminary assessment of the appropriateness of these codes have been performed for selected relevant phenomena, and application of the codes to the selected transients on the basis of defined 'reference design'. An overview on their successful upgrade to supercritical pressures and application to some plant safety analysis are provided in the paper. Further elaborations in relation to future needs are also discussed. (author)

  8. Specific features of goal setting in road traffic safety

    Science.gov (United States)

    Kolesov, V. I.; Danilov, O. F.; Petrov, A. I.

    2017-10-01

    Road traffic safety (RTS) management is inherently a branch of cybernetics and therefore requires clear formalization of the task. The paper aims at identification of the specific features of goal setting in RTS management under the system approach. The paper presents the results of cybernetic modeling of the cause-to-effect mechanism of a road traffic accident (RTA); in here, the mechanism itself is viewed as a complex system. A designed management goal function is focused on minimizing the difficulty in achieving the target goal. Optimization of the target goal has been performed using the Lagrange principle. The created working algorithms have passed the soft testing. The key role of the obtained solution in the tactical and strategic RTS management is considered. The dynamics of the management effectiveness indicator has been analyzed based on the ten-year statistics for Russia.

  9. Radiological protection. Responsibility of the Safety Engineering Company

    International Nuclear Information System (INIS)

    Netto, A.L.

    1987-01-01

    This subject takes care of the Safety Engineering at the Radiologic Protection area on the X and Gama Rays Services. It mainly emphasis the case of that companies that, due do not have proper X and Gama Rays Services utilize partime task force on this area, but answer themselves for the safety of their employees in case of any accident occurence. (author) [pt

  10. 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)

  11. Reference to the Safety Engineering Undergraduate Courses to Improve the Subjects and Contents of the Certified Safety Engineer Qualification and Examination System of China

    OpenAIRE

    Haibin Qiu; Shanghong Shi; Tingdi Zhao; Yiwei Qiao; Jiangshi Zhang

    2013-01-01

    The aim of this paper is to recommend that the subjects and contents of certified safety engineers use safety engineering undergraduate curriculum system for reference. Human resources play an important role in accident prevention and loss control. Education on safety engineering develops quickly in China. Moreover, the State Administration of Work Safety and the National Human Resources and Social Security Ministry have implemented a certified safety engineer qualification and examination sy...

  12. Patient safety - the role of human factors and systems engineering.

    Science.gov (United States)

    Carayon, Pascale; Wood, Kenneth E

    2010-01-01

    Patient safety is a global challenge that requires knowledge and skills in multiple areas, including human factors and systems engineering. In this chapter, numerous conceptual approaches and methods for analyzing, preventing and mitigating medical errors are described. Given the complexity of healthcare work systems and processes, we emphasize the need for increasing partnerships between the health sciences and human factors and systems engineering to improve patient safety. Those partnerships will be able to develop and implement the system redesigns that are necessary to improve healthcare work systems and processes for patient safety.

  13. Patient Safety: The Role of Human Factors and Systems Engineering

    Science.gov (United States)

    Carayon, Pascale; Wood, Kenneth E.

    2011-01-01

    Patient safety is a global challenge that requires knowledge and skills in multiple areas, including human factors and systems engineering. In this chapter, numerous conceptual approaches and methods for analyzing, preventing and mitigating medical errors are described. Given the complexity of healthcare work systems and processes, we emphasize the need for increasing partnerships between the health sciences and human factors and systems engineering to improve patient safety. Those partnerships will be able to develop and implement the system redesigns that are necessary to improve healthcare work systems and processes for patient safety. PMID:20543237

  14. Supervisor's experiments on radiation safety trainings in school of engineering

    International Nuclear Information System (INIS)

    Nomura, Kiyoshi

    2005-01-01

    Radiation safety training courses in School of Engineering, The University of Tokyo, were introduced. The number of radiation workers and the usage of radiation and radioisotopes have been surveyed for past 14 years. The number of radiation workers in School of Engineering has increased due to the treatment of X-ray analysis of materials, recently. It is important for workers to understand the present situation of School of Engineering before the treatment of radiation and radioisotopes. What the supervisor should tell to radiation workers were presented herewith. The basic questionnaires after the lecture are effective for radiation safety trainings. (author)

  15. Technical evaluation report on the proposed amendment to the technical specifications on the reactor protection system and the engineered safety features actuation system for Ft. Calhoun, Unit No. 1

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the application to amend the Technical Specifications for the Ft. Calhoun Unit No. 1 Nuclear Generating Plant. The review criteria are based on the Technical Specifications of St. Lucie and Calvert Cliffs, IEEE Standards, Combustion Engineering Standard Technical Specifications, and the Code of Federal Regulations. The evaluation compares the submittal made by the licensee with the NRC staff position and the review criteria and presents the reviewer's conclusion on the acceptability of the application to amend the Technical Specifications

  16. Experience with performance based training of nuclear criticality safety engineers

    International Nuclear Information System (INIS)

    Taylor, R.G.

    1993-01-01

    Historically, new entrants to the practice of nuclear criticality safety have learned their job primarily by on-the-job training (OJT) often by association with an experienced nuclear criticality safety engineer who probably also learned their job by OJT. Typically, the new entrant learned what he/she needed to know to solve a particular problem and accumulated experience as more problems were solved. It is likely that more formalism will be required in the future. Current US Department of Energy requirements for those positions which have to demonstrate qualification indicate that it should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis i's incompletely developed in some areas. Details of this analysis are provided in this report

  17. Safety-I, Safety-II and Resilience Engineering.

    Science.gov (United States)

    Patterson, Mary; Deutsch, Ellen S

    2015-12-01

    In the quest to continually improve the health care delivered to patients, it is important to understand "what went wrong," also known as Safety-I, when there are undesired outcomes, but it is also important to understand, and optimize "what went right," also known as Safety-II. The difference between Safety-I and Safety-II are philosophical as well as pragmatic. Improving health care delivery involves understanding that health care delivery is a complex adaptive system; components of that system impact, and are impacted by, the actions of other components of the system. Challenges to optimal care include regular, irregular and unexampled threats. This article addresses the dangers of brittleness and miscalibration, as well as the value of adaptive capacity and margin. These qualities can, respectively, detract from or contribute to the emergence of organizational resilience. Resilience is characterized by the ability to monitor, react, anticipate, and learn. Finally, this article celebrates the importance of humans, who make use of system capabilities and proactively mitigate the effects of system limitations to contribute to successful outcomes. Copyright © 2015 Mosby, Inc. All rights reserved.

  18. Human and organization factors: engineering operating safety into offshore structures

    International Nuclear Information System (INIS)

    Bea, Robert G.

    1998-01-01

    History indicates clearly that the safety of offshore structures is determined primarily by the humans and organizations responsible for these structures during their design, construction, operation, maintenance, and decommissioning. If the safety of offshore structures is to be preserved and improved, then attention of engineers should focus on to how to improve the reliability of the offshore structure 'system,' including the people that come into contact with the structure during its life-cycle. This article reviews and discusss concepts and engineering approaches that can be used in such efforts. Two specific human factor issues are addressed: (1) real-time management of safety during operations, and (2) development of a Safety Management Assessment System to help improve the safety of offshore structures

  19. Development of a safety case editor with assessment features

    NARCIS (Netherlands)

    Luo, Y.; Li, Z.; van den Brand, M.G.J.

    2016-01-01

    A safety case is an argumentation for showing confidence in the claimed safety assurance of a system, which should be comprehensible and well-structured. Typically, safety cases are represented in plain text, but the structure of safety cases might become ambiguous and unclear. To address this, the

  20. Standards in reliability and safety engineering

    International Nuclear Information System (INIS)

    O'Connor, Patrick

    1998-01-01

    This article explains how the highest 'world class' levels of reliability and safety are achieved, by adherence to the basic principles of excellence in design, production, support and maintenance, by continuous improvement, and by understanding that excellence and improvement lead to reduced costs. These principles are contrasted with the methods that have been developed and standardised, particularly military standards for reliability, ISO9000, and safety case regulations. The article concludes that the formal, standardised approaches are misleading and counterproductive, and recommends that they be replaced by a philosophy based on the realities of human performance

  1. Safety and maintenance engineering: A compilation

    Science.gov (United States)

    1974-01-01

    A compilation is presented for the dissemination of information on technological developments which have potential utility outside the aerospace and nuclear communities. Safety of personnel engaged in the handling of hazardous materials and equipment, protection of equipment from fire, high wind, or careless handling by personnel, and techniques for the maintenance of operating equipment are reported.

  2. 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.

  3. 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

  4. The science and engineering of HIFAR safety

    International Nuclear Information System (INIS)

    Connolly, J.W.; Clancy, B.E.; Beathie, D.R.H.; Robinson, G.S.; Godfrey, R.M.; Harrington, B.V.

    1993-12-01

    Since the HIFAR Safety Document was first issued, major improvements have occurred in the quality of data and in the methods of calculation which are available for deterministic analysis of the behaviour of the reactor in normal or in accident conditions. Many such analyses have been carried out but the results have been reported in a wide range of internal memoranda and in external reports. In this report the most significant of the improved methods are described and the results of some of those analyses are reviewed. Principal areas covered are reactor physics of the core and reflector, the dynamics of the control systems, thermal hydraulic aspects important to safety margins, and the emergency core cooling system. Abnormal events discussed are inadvertent reactivity insertion sequences and the loss of coolant accident. Where possible, consistent sets of data are provided for use in future analyses. 35 refs., 23 tabs., 28 figs

  5. Should Nuclear Safety Care About Resilience Engineering?

    International Nuclear Information System (INIS)

    Paries, J.

    2016-01-01

    The current nuclear industry safety paradigm is based on the deterministic and/or probabilistic anticipation of all potential situations, and the predetermination of all the (safe) responses. Even the defense in depth concept, which is the core of the nuclear safety strategy and is intended to handle situations in which part of the control is lost, heavily relies on detailed anticipations. In other words, nuclear safety is mainly expected from the real world’s conformity to a designed-to-be-safe world, i.e., a well controlled world, where organizations, processes, hardware, teams, and individuals comply with their rationally predetermined behaviors. In this “command and control” perspective, risk is seen as mainly generated by deviations and variations from rules, procedures, norms, and expectations. However, real operations are complex, even in normal situations, which means that they include some unpredictable events and adaptation behaviors. The traditional “command and control” perspective fail to properly acknowledge the limits to predictability inherent to a complex adaptive system. It actually strives to reduce complexity through tighter compliance to specifications and to improve predictions capabilities through a tighter monitoring of “weak signals” and “precursors”. But in a complex world, precursors are usually obvious after the event, while not identifiable before. And the efforts made to reduce complexity may also simultaneously tighten couplings between system’s components— hence increase complexity — and reduce the diversity and flexibility needed to respond to it.

  6. Safety in offshore engineering an academic course covering safety in offshore wind

    NARCIS (Netherlands)

    Cerda Salzmann, D.J.

    2011-01-01

    Offshore projects are known for their challenging conditions, generally leading to high risks. Therefore no offshore project can go without a continuous and extensive assessment on safety issues. The Delft University of Technology is currently developing a course "Safety in Offshore Engineering"

  7. Fire-safety engineering and performance-based codes

    DEFF Research Database (Denmark)

    Sørensen, Lars Schiøtt

    project administrators, etc. The book deals with the following topics: • Historical presentation on the subject of fire • Legislation and building project administration • European fire standardization • Passive and active fire protection • Performance-based Codes • Fire-safety Engineering • Fundamental......Fire-safety Engineering is written as a textbook for Engineering students at universities and other institutions of higher education that teach in the area of fire. The book can also be used as a work of reference for consulting engineers, Building product manufacturers, contractors, building...... thermodynamics • Heat exchange during the fire process • Skin burns • Burning rate, energy release rate and design fires • Proposal to Risk-based design fires • Proposal to a Fire scale • Material ignition and flame spread • Fire dynamics in buildings • Combustion products and toxic gases • Smoke inhalation...

  8. Product Engineering Class in the Software Safety Risk Taxonomy for Building Safety-Critical Systems

    Science.gov (United States)

    Hill, Janice; Victor, Daniel

    2008-01-01

    When software safety requirements are imposed on legacy safety-critical systems, retrospective safety cases need to be formulated as part of recertifying the systems for further use and risks must be documented and managed to give confidence for reusing the systems. The SEJ Software Development Risk Taxonomy [4] focuses on general software development issues. It does not, however, cover all the safety risks. The Software Safety Risk Taxonomy [8] was developed which provides a construct for eliciting and categorizing software safety risks in a straightforward manner. In this paper, we present extended work on the taxonomy for safety that incorporates the additional issues inherent in the development and maintenance of safety-critical systems with software. An instrument called a Software Safety Risk Taxonomy Based Questionnaire (TBQ) is generated containing questions addressing each safety attribute in the Software Safety Risk Taxonomy. Software safety risks are surfaced using the new TBQ and then analyzed. In this paper we give the definitions for the specialized Product Engineering Class within the Software Safety Risk Taxonomy. At the end of the paper, we present the tool known as the 'Legacy Systems Risk Database Tool' that is used to collect and analyze the data required to show traceability to a particular safety standard

  9. Nuclear reactor safety: physics and engineering aspects

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1982-01-01

    In order to carry out the sort of probabilistic analysis referred to by Farmer (Contemp. Phys.; 22:349(1981)), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in sections dealing with the neutron chain reaction, the removal of heat from the reactor, material problems, reliability of protective systems and a number of specific topics of particular interest from the point of view of physics or engineering. (author)

  10. Inherent Safety Feature of Hybrid Low Power Research Reactor during Reactivity Induced Accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, DongHyun; Yum, Soo Been; Hong, Sung Teak; Lim, In-Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hybrid low power research reactor(H-LPRR) is the new design concept of low power research reactor for critical facility as well as education and training. In the case of typical low power research reactor, the purposes of utilization are the experiments for education of nuclear engineering students, Neutron Activation Analysis(NAA) and radio-isotope production for research purpose. H-LPRR is a light-water cooled and moderated research reactor that uses rod-type LEU UO{sub 2} fuels same as those for commercial power plants. The maximum core thermal power is 70kW and, the core is placed in the bottom of open pool. There are 1 control rod and 2 shutdown rods in the core. It is designed to cool the core by natural convection, retain negative feedback coefficient for entire fuel periods and operate for 20 years without refueling. Inherent safety in H-LPRR is achieved by passive design features such as negative temperature feedback coefficient and core cooling by natural convection during normal and emergency conditions. The purpose of this study is to find out that the inherent safety characteristics of H-LPRR is able to control the power and protect the reactor from the RIA(Reactivity induced accident). RIA analysis was performed to investigate the inherent safety feature of H-LPRR. As a result, it was found that the reactor controls its power without fuel damage in the event and that the reactor remains safe states inherently. Therefore, it is believed that high degree of safety inheres in H-LPRR.

  11. Temperature calculation in fire safety engineering

    CERN Document Server

    Wickström, Ulf

    2016-01-01

    This book provides a consistent scientific background to engineering calculation methods applicable to analyses of materials reaction-to-fire, as well as fire resistance of structures. Several new and unique formulas and diagrams which facilitate calculations are presented. It focuses on problems involving high temperature conditions and, in particular, defines boundary conditions in a suitable way for calculations. A large portion of the book is devoted to boundary conditions and measurements of thermal exposure by radiation and convection. The concepts and theories of adiabatic surface temperature and measurements of temperature with plate thermometers are thoroughly explained. Also presented is a renewed method for modeling compartment fires, with the resulting simple and accurate prediction tools for both pre- and post-flashover fires. The final chapters deal with temperature calculations in steel, concrete and timber structures exposed to standard time-temperature fire curves. Useful temperature calculat...

  12. Requirements Engineering for Software Integrity and Safety

    Science.gov (United States)

    Leveson, Nancy G.

    2002-01-01

    Requirements flaws are the most common cause of errors and software-related accidents in operational software. Most aerospace firms list requirements as one of their most important outstanding software development problems and all of the recent, NASA spacecraft losses related to software (including the highly publicized Mars Program failures) can be traced to requirements flaws. In light of these facts, it is surprising that relatively little research is devoted to requirements in contrast with other software engineering topics. The research proposed built on our previous work. including both criteria for determining whether a requirements specification is acceptably complete and a new approach to structuring system specifications called Intent Specifications. This grant was to fund basic research on how these ideas could be extended to leverage innovative approaches to the problems of (1) reducing the impact of changing requirements, (2) finding requirements specification flaws early through formal and informal analysis, and (3) avoiding common flaws entirely through appropriate requirements specification language design.

  13. Rasmussen's legacy: A paradigm change in engineering for safety.

    Science.gov (United States)

    Leveson, Nancy G

    2017-03-01

    This paper describes three applications of Rasmussen's idea to systems engineering practice. The first is the application of the abstraction hierarchy to engineering specifications, particularly requirements specification. The second is the use of Rasmussen's ideas in safety modeling and analysis to create a new, more powerful type of accident causation model that extends traditional models to better handle human-operated, software-intensive, sociotechnical systems. Because this new model has a formal, mathematical foundation built on systems theory (as was Rasmussen's original model), new modeling and analysis tools become possible. The third application is to engineering hazard analysis. Engineers have traditionally either omitted human from consideration in system hazard analysis or have treated them rather superficially, for example, that they behave randomly. Applying Rasmussen's model of human error to a powerful new hazard analysis technique allows human behavior to be included in engineering hazard analysis. Copyright © 2016 Elsevier Ltd. All rights reserved.

  14. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  15. New engineering safety factors for Loviisa NPP core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuopanportti, Jaakko; Saarinen, Simo; Lahtinen, Tuukka; Ekstroem, Karoliina [Fortum Power and Heat Ltd., Fortum (Finland)

    2017-09-15

    In Loviisa NPP, there are two limiting thermal margins called the enthalpy rise margin and the linear heat rate margin that are monitored during normal operation. Engineering safety factors are applied in determination of both of these factors. The factors take into account the effect of various manufacturing tolerances, impact of the irradiation and simulation uncertainties on the local heat rate and on the enthalpy of the coolant. The engineering factors were re-evaluated during 2015 and the factors were approved by the Finnish radiation and nuclear safety authority in 2016. The re-evaluation was performed by considering all of the identified phenomena that affect the local heat rate or the enthalpy of the coolant. This paper summarizes the work that was performed during the re-evaluation of the engineering safety factors and presents the results for each uncertainty component. The new engineering safety factors are 1.115 for the linear heat rate and 1.100 for the enthalpy rise margin when the old factors were 1.12 and 1.16, respectively. The new factors improve the fuel economy by about 1%.

  16. Legal bases of safety regulations in electrical engineering

    Energy Technology Data Exchange (ETDEWEB)

    Jeiter, W

    1981-12-01

    Apart from the governmental regulations the rule for the prevention of accidents 'Electric plants and equipment' must be observed in order to protect the insurants. Actually, all these regulations do not contain any independent instructions. They rather utilize the VDE regulations and refer to them. The laws of electrical safety engineering are strongly influenced by harmonization efforts particularly within the European Communitties.

  17. CANDU combined cycles featuring gas-turbine engines

    International Nuclear Information System (INIS)

    Vecchiarelli, J.; Choy, E.; Peryoga, Y.; Aryono, N.A.

    1998-01-01

    In the present study, a power-plant analysis is conducted to evaluate the thermodynamic merit of various CANDU combined cycles in which continuously operating gas-turbine engines are employed as a source of class IV power restoration. It is proposed to utilize gas turbines in future CANDU power plants, for sites (such as Indonesia) where natural gas or other combustible fuels are abundant. The primary objective is to eliminate the standby diesel-generators (which serve as a backup supply of class III power) since they are nonproductive and expensive. In the proposed concept, the gas turbines would: (1) normally operate on a continuous basis and (2) serve as a reliable backup supply of class IV power (the Gentilly-2 nuclear power plant uses standby gas turbines for this purpose). The backup class IV power enables the plant to operate in poison-prevent mode until normal class IV power is restored. This feature is particularly beneficial to countries with relatively small and less stable grids. Thermodynamically, the advantage of the proposed concept is twofold. Firstly, the operation of the gas-turbine engines would directly increase the net (electrical) power output and the overall thermal efficiency of a CANDU power plant. Secondly, the hot exhaust gases from the gas turbines could be employed to heat water in the CANDU Balance Of Plant (BOP) and therefore improve the thermodynamic performance of the BOP. This may be accomplished via several different combined-cycle configurations, with no impact on the current CANDU Nuclear Steam Supply System (NSSS) full-power operating conditions when each gas turbine is at maximum power. For instance, the hot exhaust gases may be employed for feedwater preheating and steam reheating and/or superheating; heat exchange could be accomplished in a heat recovery steam generator, as in conventional gas-turbine combined-cycle plants. The commercially available GateCycle power plant analysis program was applied to conduct a

  18. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  19. Determination of engineering safety factor -routine in Hungary (a methodology for the normal operation local power engineering safety factors)

    International Nuclear Information System (INIS)

    Szecsenyi, Z.; Korpas, L.; Bona, G.; Kereszturi, A.

    2010-01-01

    From the late nineties Paks Nuclear Power Plant-in collaboration with KFKI Atomic Energy Research Institute (KFKI AEKI)- is developing a system for determining the normal operation local power engineering safety factors. The system is based on a Monte Carlo sampling of the uncertain model input parameters. Additionally, the comparison of the calculation to the in-core measurements plays essential role for determining some important input parameters. By using new fuel types and the corresponding more recent detailed technological data, the applied method is being improved from time to time. Presently, the actually used and authorized engineering safety factors at Paks NPP are determined by using this method. In the paper, the system.s main properties are described (not going beyond the possible extent). The main points are as follows:-Mathematical definition of the engineering safety factor;-Sources of the uncertainties;-Input error propagation method constituting the basis of the system;-Flow-chart of the subsequent steps of the determination Finally, in the paper the engineering safety factors values of some selected parameters are presented as examples for demonstration of the capability of the method. (Authors)

  20. 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

  1. 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.

  2. TELEPERM XS: I and C systems for safety application in NPP's - features, developments, references and feedback

    International Nuclear Information System (INIS)

    Prehler, Heinz Josef

    2007-01-01

    In the field of digital I and C AREVA NP is focused on concepts that on the one hand make allowance for development cycles getting shorter in the technology competition and on the other hand assure a long-term system support with the ability to deliver spare parts in the long run. The system platform TELEPERM XS, which was developed especially for safety I and C application of nuclear power plants, meets requirements effectively and thus provides a great benefit for the customer. The typical applications of TELEPERM XS are in the field of Reactor Protection and ESFAS functions (Engineered Safety Features Actuation System). High demands are defined for system reliability and availability, as well as for failure prevention and tolerance. The requirements of corresponding international codes and standards of nuclear installations are also implemented in the development and engineering processes of TELEPERM XS. The system platform is integrated into a sustainable program for service life management of electronic systems and equipment. Its ongoing future-oriented development ensures the long-term availability of hardware and software components for installed TELEPERM XS applications already installed in the plants. The further development of platform and components continues to be based on the robust, service-proven TELEPERM XS architecture, with the aim of minimizing the risks associated with equipment qualification and project licensing. A further development feature is the completion and extension of TELEPERM XS applications. This continuous innovation process, combined with maximized compatibility, makes TELEPERM XS unique, and provides the basis for a sustainable system with a service life guaranteed for the long term. Within the past 10 years, the majority of all comprehensive modernization projects worldwide were implemented or are contracted using TELEPERM XS. TELEPERM XS has been implemented in two new nuclear power plants and there are orders for four more

  3. Transport fire safety engineering in the European Union - project TRANSFEU

    Directory of Open Access Journals (Sweden)

    Jolanta Maria RADZISZEWSKA-WOLIŃSKA

    2011-01-01

    Full Text Available Article presents European Research project (of FP7-SST-2008-RTD-1 for Surface transportation TRANSFEU. Projects undertakes to deliver both a reliable toxicity measurement methodology and a holistic fire safety approach for all kind of surface transport. It bases on a harmonized Fire Safety Engineering methodology which link passive fire security with active fire security mode. This all embracing system is the key to attain optimum design solutions in respect to fire safety objectives as an alternative to the prescriptive approach. It will help in the development of innovative solutions (design and products used for the building of the surface transport which will better respect the environment.In order to reach these objectives new toxicity measurement methodology and related classification of materials, new numerical fire simulation tools, fire test methodology (laboratory and full scale and a decisive tool to optimize or explore new design in accordance to the fire safety requirements will be developed.

  4. 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

  5. Image processing for safety assessment in civil engineering.

    Science.gov (United States)

    Ferrer, Belen; Pomares, Juan C; Irles, Ramon; Espinosa, Julian; Mas, David

    2013-06-20

    Behavior analysis of construction safety systems is of fundamental importance to avoid accidental injuries. Traditionally, measurements of dynamic actions in civil engineering have been done through accelerometers, but high-speed cameras and image processing techniques can play an important role in this area. Here, we propose using morphological image filtering and Hough transform on high-speed video sequence as tools for dynamic measurements on that field. The presented method is applied to obtain the trajectory and acceleration of a cylindrical ballast falling from a building and trapped by a thread net. Results show that safety recommendations given in construction codes can be potentially dangerous for workers.

  6. Health and safety implications of occupational exposure to engineered nanomaterials.

    Science.gov (United States)

    Stebounova, Larissa V; Morgan, Hallie; Grassian, Vicki H; Brenner, Sara

    2012-01-01

    The rapid growth and commercialization of nanotechnology are currently outpacing health and safety recommendations for engineered nanomaterials. As the production and use of nanomaterials increase, so does the possibility that there will be exposure of workers and the public to these materials. This review provides a summary of current research and regulatory efforts related to occupational exposure and medical surveillance for the nanotechnology workforce, focusing on the most prevalent industrial nanomaterials currently moving through the research, development, and manufacturing pipelines. Their applications and usage precedes a discussion of occupational health and safety efforts, including exposure assessment, occupational health surveillance, and regulatory considerations for these nanomaterials. Copyright © 2011 Wiley Periodicals, Inc.

  7. Assessment of safety engineering of circuits with dc micromotors

    Energy Technology Data Exchange (ETDEWEB)

    Pavlyuchenko, L.A.; Starchuk, S.E.

    1986-01-01

    Presents an assessment of safety engineering in d.c. micromotors operating as part of actuating devices in mining equipment. These micromotors should have RO (especially explosion proof) protection. The safety engineering should be assessed with an intermittent fault in the power line. Equations are given for calculation of the equivalent inductance of the micromotor circuit with an intermittent power line fault. If the circuit is not intrinsically safe, a diode in the forward direction is recommended for connection in series with the micromotor. If the power line is not intrinsically safe, a diode shunt is recommended. Comparative data for power sources (IBP) and micromotors (DPM, DPR, with permanent magnets) are given in tables. 4 refs.

  8. A survey of characteristic engine features for technology-sustained pervasive games

    CERN Document Server

    Nevelsteen, Kim JL

    2015-01-01

    This book scrutinizes pervasive games from a technological perspective, focusing on the sub-domain of games that satisfy the criteria that they make use of virtual game elements. In the computer game industry, the use of a game engine to build games is common, but current game engines do not support pervasive games. Since the computer game industry is already rich with game engines, this book investigates: (i) if a game engine can be repurposed to stage pervasive games; (ii) if features describing a would-be pervasive game engine can be identified; (iii) using those features, if an architectu

  9. Inherent safety features of the HTTR revealed in the accident condition

    International Nuclear Information System (INIS)

    Kunitomi, K.; Shinozaki, M.; Baba, O.; Saito, S.

    1992-01-01

    The High Temperature Engineering Test Reactor (HTTR) being constructed by JAERI (Japan Atomic Energy Research Institute) is a graphite-moderated and helium-cooled reactor with an outlet gas temperature of 950degC. The inherent safety characteristics in the HTTR prevent temperature increase of reactor fuels and fission product release from the reactor core in postulated accident conditions. The reactor core can be cooled by a Vessel Cooling System (VCS) indirectly, even in the case that no forced cooling is expected during the accident such as primary pipe break. The VCS consists of independent water cooling loop and cooling panel around the reactor pressure vessel. The cooling panel whose temperature of 60-90degC cools the reactor pressure vessel by radiation and removes the decay heat from the core indirectly. Furthermore, even if failure of VCS is assumed during this accident as a severe accident, the reactor core is remained safe despite the temperature increase of biological concrete shield around the reactor pressure vessel. This paper describes the inherent safety features of the HTTR specially focused on the accident condition without forced cooling. The detailed analytical results of such an accident are described together with clarifying the role of the VCS. (author)

  10. Safety considerations in the design of the fusion engineering device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    Safety considerations play a significant role in the design of a near-term Fusion Engineering Device (FED). For the safety of the general public and the plant workers, the radiation environment caused by the reacting plasma and the potential release of tritium fuel are the dominant considerations. The U.S. Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris. The highly activated device components make it necessary to design many of the maintenance activities in the reactor building for totally remote operation. The hot cell facility has evolved as a totally remote maintenance facility due to the high radiation levels of the device components. Safety considerations have had substantial impacts on the design of FED. Several examples of safety-related design impacts are discussed in the paper. Feasible solutions have been identified for all outstanding safety-related items, and additional optimization of these solutions is anticipated in future design studies

  11. Engineering approach to relative quantitative assessment of safety culture and related social issues in NPP operation

    International Nuclear Information System (INIS)

    Sivokon, V.; Gladyshev, M.; Malkin, S.

    2005-01-01

    The report is devoted to presentation of engineering approach and software tool developed for Safety Culture (SC) assessment as well as to the results of their implementation at Smolensk NPP. The engineering approach is logic evolution of the IAEA ASSET method broadly used at European NPPs in 90-s. It was implemented at Russian and other plants including Olkiluoto NPP in Finland. The approach allows relative quantitative assessing and trending the aspects of SC by the analysis of evens features and causes, calculation and trending corresponding indicators. At the same time plant's operational performances and related social issues, including efficiency of plant operation and personnel reliability, can be monitored. With the help of developed tool the joint team combined from personnel of Smolensk NPP and RRC 'Kurchatov Institute' ('KI') issued the SC self-assessment report, which identifies: families of recurrent events, main safety and operational problems ; their trends and importance to SC and plant efficiency; recommendations to enhance SC and operational performance

  12. Engineering thinking in emergency situations: A new nuclear safety concept.

    Science.gov (United States)

    Guarnieri, Franck; Travadel, Sébastien

    2014-11-01

    The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for "engineering thinking in emergency situations." This is a new concept that emphasizes adaptability and resilience within organizations-such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclear safety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident.

  13. Design of 3D simulation engine for oilfield safety training

    Science.gov (United States)

    Li, Hua-Ming; Kang, Bao-Sheng

    2015-03-01

    Aiming at the demand for rapid custom development of 3D simulation system for oilfield safety training, this paper designs and implements a 3D simulation engine based on script-driven method, multi-layer structure, pre-defined entity objects and high-level tools such as scene editor, script editor, program loader. A scripting language been defined to control the system's progress, events and operating results. Training teacher can use this engine to edit 3D virtual scenes, set the properties of entity objects, define the logic script of task, and produce a 3D simulation training system without any skills of programming. Through expanding entity class, this engine can be quickly applied to other virtual training areas.

  14. The function of specialized organization in work safety engineering for nuclear installations

    International Nuclear Information System (INIS)

    Salvatore, J.E.L.

    1989-01-01

    The attributions of Brazilian CNEN in the licensing procedures of any nuclear installation are discussed. It is shown that the work safety engineering and industrial safety constitute important functions for nuclear safety. (M.C.K.) [pt

  15. Engineering judgement and bridging the fire safety gap in existing nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Qamheiah, G.; Wu, Y., E-mail: gqamheiah@plcfire.com, E-mail: dwu@plcfire.com [PLC Fire Safety Solutions, Mississauga, ON (Canada)

    2014-07-01

    Canadian nuclear power plants were constructed in the 1960's through the 1980's. Fire safety considerations were largely based on guidance from general building and fire codes in effect at the time. Since then, nuclear specific fire safety standards have been developed and adopted by the Regulator, increasing the expected level of fire safety in the process. Application of the standards to existing plants was largely limited to operational requirements viewed as retroactive. However, as existing facilities undergo modifications or refurbishment for the purpose of life extension, the expectation is that the design requirements of these fire safety standards also be satisfied. This creates considerable challenges for existing nuclear power plants as fire safety requirements such as those intended to assure means for safe egress, prevention of fire spread and protection of redundancy rely upon fire protection features that are inherent in the physical infrastructural design. This paper focuses on the methodology for conducting fire safety gap analyses on existing plants, and the integral role that engineering judgement plays in the development of viable and cost effective solutions to achieve the objectives of the current fire safety standards. (author)

  16. FEATURES TERMINOLOGY IN MODERN MEDICAL SCIENCE AND ENGINEERING

    Directory of Open Access Journals (Sweden)

    Zlepko S.M.

    2016-02-01

    Full Text Available The article is devoted to the problem of compliance with terms and definitions in medical science and engineering to the actual essence. One of the components of successful development of these trends is adequate linguistic support of the process of development and operation, basic level of determination and terms which indicated certain principles, approaches, processes and so on.

  17. The design features and safety concepts of the nuclear heating reactor developed in China

    International Nuclear Information System (INIS)

    Zheng Wenxiang; Wang Dazhong

    1995-01-01

    Based on the specific conditions of the nuclear heat applications and the development objectives of the advanced reactors, the nuclear heating reactor (NHR) exploited in China has adhered to the new safety concepts and been designed with a number of advanced features, including the integrated arrangement, full power natural circulation capacity, self-pressurized performance, dynamically-hydraulic control rod drive and passive safety systems, so that higher standard of safety as well as simplification in the plant systems and improvement in economic viability has been achieved. This paper describes the special consideration in the design as well as the main design features and safety concepts of the NHR. Some experimental and analytical results are also presented to demonstrate the NHR safety features

  18. A software engineering process for safety-critical software application

    International Nuclear Information System (INIS)

    Kang, Byung Heon; Kim, Hang Bae; Chang, Hoon Seon; Jeon, Jong Sun

    1995-01-01

    Application of computer software to safety-critical systems in on the increase. To be successful, the software must be designed and constructed to meet the functional and performance requirements of the system. For safety reason, the software must be demonstrated not only to meet these requirements, but also to operate safely as a component within the system. For longer-term cost consideration, the software must be designed and structured to ease future maintenance and modifications. This paper presents a software engineering process for the production of safety-critical software for a nuclear power plant. The presentation is expository in nature of a viable high quality safety-critical software development. It is based on the ideas of a rational design process and on the experience of the adaptation of such process in the production of the safety-critical software for the shutdown system number two of Wolsung 2, 3 and 4 nuclear power generation plants. This process is significantly different from a conventional process in terms of rigorous software development phases and software design techniques, The process covers documentation, design, verification and testing using mathematically precise notations and highly reviewable tabular format to specify software requirements and software requirements and software requirements and code against software design using static analysis. The software engineering process described in this paper applies the principle of information-hiding decomposition in software design using a modular design technique so that when a change is required or an error is detected, the affected scope can be readily and confidently located. it also facilitates a sense of high degree of confidence in the 'correctness' of the software production, and provides a relatively simple and straightforward code implementation effort. 1 figs., 10 refs. (Author)

  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. 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

  1. The Alternative Design Features for Safety Enhancement in Shutdown Operation

    International Nuclear Information System (INIS)

    Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro

    2009-01-01

    PSA can be used to confirm that the new plant design is complied with the applicable safety goals, and to select among the alternate design options. A shutdown PSA provides insight for outage planning schedule, outage management practices, and design modifications. Considering the results of both LPSD PSA studies and operating experiences for low power and shutdown, the improvements can be proposed to reduce the high risk contribution. The improvements/enhancements during shutdown operation may be divided into categories such as hardware, administrative management, and operational procedure. This paper presents on an example how the risk related to an accidental situation can be reduced, focusing the hardware design changes for the newly designed NPPs

  2. Technical and institutional safety features of nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Rosa, L.P.

    1986-01-01

    This work reports technical, political and institutional safety features of nuclear power plants in Brazil. It is mainly concerned with reactor accidents and personnel safety. The three mile Island and Chernobyl accidents are also discussed and taken as examples. (A.C.A.S.)

  3. Feature-based engineering of compensations in web service environment

    DEFF Research Database (Denmark)

    Schaefer, Michael; Dolog, Peter

    2009-01-01

    In this paper, we introduce a product line approach for developing Web services with extended compensation capabilities. We adopt a feature modelling approach in order to describe variable and common compensation properties of Web service variants, as well as service consumer application...

  4. Application of software engineering to development of reactor safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1981-01-01

    Software Engineering, which is a systematic methodology by which a large scale software development project is partitioned into manageable pieces, has been applied to the development of LMFBR safety codes. The techniques have been applied extensively in the business and aerospace communities and have provided an answer to the drastically increasing cost of developing and maintaining software. The five phases of software engineering (Survey, Analysis, Design, Implementation, and Testing) were applied in turn to development of these codes, along with Walkthroughs (peer review) at each stage. The application of these techniques has resulted in SUPERIOR SOFTWARE which is well documented, thoroughly tested, easy to modify, easier to use and maintain. The development projects have resulted in lower overall cost. (orig.) [de

  5. Reliability and Maintainability Engineering - A Major Driver for Safety and Affordability

    Science.gov (United States)

    Safie, Fayssal M.

    2011-01-01

    The United States National Aeronautics and Space Administration (NASA) is in the midst of an effort to design and build a safe and affordable heavy lift vehicle to go to the moon and beyond. To achieve that, NASA is seeking more innovative and efficient approaches to reduce cost while maintaining an acceptable level of safety and mission success. One area that has the potential to contribute significantly to achieving NASA safety and affordability goals is Reliability and Maintainability (R&M) engineering. Inadequate reliability or failure of critical safety items may directly jeopardize the safety of the user(s) and result in a loss of life. Inadequate reliability of equipment may directly jeopardize mission success. Systems designed to be more reliable (fewer failures) and maintainable (fewer resources needed) can lower the total life cycle cost. The Department of Defense (DOD) and industry experience has shown that optimized and adequate levels of R&M are critical for achieving a high level of safety and mission success, and low sustainment cost. Also, lessons learned from the Space Shuttle program clearly demonstrated the importance of R&M engineering in designing and operating safe and affordable launch systems. The Challenger and Columbia accidents are examples of the severe impact of design unreliability and process induced failures on system safety and mission success. These accidents demonstrated the criticality of reliability engineering in understanding component failure mechanisms and integrated system failures across the system elements interfaces. Experience from the shuttle program also shows that insufficient Reliability, Maintainability, and Supportability (RMS) engineering analyses upfront in the design phase can significantly increase the sustainment cost and, thereby, the total life cycle cost. Emphasis on RMS during the design phase is critical for identifying the design features and characteristics needed for time efficient processing

  6. EPRTM engineered features for core melt mitigation in severe accidents

    International Nuclear Information System (INIS)

    Fischer, Manfred; Henning, Andreas

    2009-01-01

    For the prevention of accident conditions, the EPR TM relies on the proven 3-level safety concepts inherited from its predecessors, the French 'N4' and the German 'Konvoi' NPP. In addition, a new, fourth 'beyond safety' level is implemented for the mitigation of postulated severe accidents (SA) with core melting. It is aimed at preserving the integrity of the containment barrier and at significantly reducing the frequency and magnitude of activity releases into the environment under such extreme conditions. Loss of containment integrity is prevented by dedicated design measures that address short- and long-term challenges, like: the melt-through of the reactor pressure vessel under high internal pressure, energetic hydrogen/steam explosions, containment overpressure failure, and basemat melt-through. The EPR TM SA systems and components that address these issues are: - the dedicated SA valves for the depressurization the primary circuit, - the provisions for H 2 recombination, atmospheric mixing, steam dilution, - the core melt stabilization system, - the dedicated SA containment heat removal system. The core melt stabilization system (CMSS) of the EPR TM is based on a two-stage ex-vessel approach. After its release from the RPV the core debris is first accumulated and conditioned in the (dry) reactor pit by the addition of sacrificial concrete. Then the created molten pool is spread into a lateral core catcher to establish favorable conditions for the later flooding, quenching and cooling with water passively drained from the Internal Refueling Water Storage Tank. Long-term heat removal from the containment is achieved by sprays that are supplied with water by the containment heat removal system. Complementing earlier publications focused on the principle function, basic design, and validation background of the EPR TM CMSS, this paper describes the state achieved after detailed design, as well as the technical solutions chosen for its main components, including

  7. Improvement of Safety Features in Standard Operation Procedure of Tc-99m Generator

    International Nuclear Information System (INIS)

    Manisah Saedon; Mohd Khairul Hakimi; Shyen, A.K.S.

    2011-01-01

    This paper describes the improvements proposed to the original production procedures for Tc-99m generators. Improvements are intended to add safety and health features for workers into the existing procedures. The difference between the new safe work procedures from the original work procedures; is the concern about the safety and health of employees other than the product safety. One of the suggested safety characteristics is by using the visual aid so that the workers can easily see and read the procedures when they perform their duties, whereas the previous procedures are kept in the manual and difficult to access. The purpose of this paper is to share information about the importance of safety and health features for the workers in the procedures established in addition to provide awareness to all parties involved. (author)

  8. The Use of a Parametric Feature Based CAD System to Teach Introductory Engineering Graphics.

    Science.gov (United States)

    Howell, Steven K.

    1995-01-01

    Describes the use of a parametric-feature-based computer-aided design (CAD) System, AutoCAD Designer, in teaching concepts of three dimensional geometrical modeling and design. Allows engineering graphics to go beyond the role of documentation and communication and allows an engineer to actually build a virtual prototype of a design idea and…

  9. 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

  10. The role of engineering judgement, safety culture, and organizational factors in risk assessment

    International Nuclear Information System (INIS)

    Muzumdar, Ajit; Professor, Visiting

    1996-01-01

    This paper reviews the role of engineering judgement, safety culture, and organizational factors in risk assessment by examining the reasons for human-based error. The need for more emphasis on producing engineers with good engineering judgement is described. The progress in quantifying the role of safety culture and organizational factors in risk assessment studies is summarized

  11. Topological features of engineered arrays of adsorbates in honeycomb lattices

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Arraga, Luis A., E-mail: ludovici83@gmail.com [IMDEA Nanociencia, Calle de Faraday, 9, Cantoblanco, 28049 Madrid (Spain); Lado, J.L. [International Iberian Nanotechnology Laboratory (INL), Av. Mestre Jose Veiga, 4715-330 Braga (Portugal); Guinea, Francisco [IMDEA Nanociencia, Calle de Faraday, 9, Cantoblanco, 28049 Madrid (Spain); School of Physics and Astronomy, University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom)

    2016-09-01

    Hydrogen adatoms are one of the most the promising proposals for the functionalization of graphene. The adatoms induce narrow resonances near the Dirac energy, which lead to the formation of magnetic moments. Furthermore, they also create local lattice distortions which enhance the spin–orbit coupling. The combination of magnetism and spin–orbit coupling allows for a rich variety of phases, some of which have non-trivial topological features. We analyze the interplay between magnetism and spin–orbit coupling in ordered arrays of adsorbates on honeycomb lattice monolayers, and classify the different phases that may arise. We extend our model to consider arrays of adsorbates in graphene-like crystals with stronger intrinsic spin–orbit couplings. We also consider a regime away from half-filling in which the Fermi level is at the bottom of the conduction band, we find a Berry curvature distribution corresponding to a Valley–Hall effect.

  12. Engineering and Safety Partnership Enhances Safety of the Space Shuttle Program (SSP)

    Science.gov (United States)

    Duarte, Alberto

    2007-01-01

    Project Management must use the risk assessment documents (RADs) as tools to support their decision making process. Therefore, these documents have to be initiated, developed, and evolved parallel to the life of the project. Technical preparation and safety compliance of these documents require a great deal of resources. Updating these documents after-the-fact not only requires substantial increase in resources - Project Cost -, but this task is also not useful and perhaps an unnecessary expense. Hazard Reports (HRs), Failure Modes and Effects Analysis (FMEAs), Critical Item Lists (CILs), Risk Management process are, among others, within this category. A positive action resulting from a strong partnership between interested parties is one way to get these documents and related processes and requirements, released and updated in useful time. The Space Shuttle Program (SSP) at the Marshall Space Flight Center has implemented a process which is having positive results and gaining acceptance within the Agency. A hybrid Panel, with equal interest and responsibilities for the two larger organizations, Safety and Engineering, is the focal point of this process. Called the Marshall Safety and Engineering Review Panel (MSERP), its charter (Space Shuttle Program Directive 110 F, April 15, 2005), and its Operating Control Plan emphasizes the technical and safety responsibilities over the program risk documents: HRs; FMEA/CILs; Engineering Changes; anomalies/problem resolutions and corrective action implementations, and trend analysis. The MSERP has undertaken its responsibilities with objectivity, assertiveness, dedication, has operated with focus, and has shown significant results and promising perspectives. The MSERP has been deeply involved in propulsion systems and integration, real time technical issues and other relevant reviews, since its conception. These activities have transformed the propulsion MSERP in a truly participative and value added panel, making a

  13. Engineering nanomaterials-based biosensors for food safety detection.

    Science.gov (United States)

    Lv, Man; Liu, Yang; Geng, Jinhui; Kou, Xiaohong; Xin, Zhihong; Yang, Dayong

    2018-05-30

    Food safety always remains a grand global challenge to human health, especially in developing countries. To solve food safety pertained problems, numerous strategies have been developed to detect biological and chemical contaminants in food. Among these approaches, nanomaterials-based biosensors provide opportunity to realize rapid, sensitive, efficient and portable detection, overcoming the restrictions and limitations of traditional methods such as complicated sample pretreatment, long detection time, and relying on expensive instruments and well-trained personnel. In this review article, we provide a cross-disciplinary perspective to review the progress of nanomaterials-based biosensors for the detection of food contaminants. The review article is organized by the category of food contaminants including pathogens/toxins, heavy metals, pesticides, veterinary drugs and illegal additives. In each category of food contaminant, the biosensing strategies are summarized including optical, colorimetric, fluorescent, electrochemical, and immune- biosensors; the relevant analytes, nanomaterials and biosensors are analyzed comprehensively. Future perspectives and challenges are also discussed briefly. We envision that our review could bridge the gap between the fields of food science and nanotechnology, providing implications for the scientists or engineers in both areas to collaborate and promote the development of nanomaterials-based biosensors for food safety detection. Copyright © 2018 Elsevier B.V. All rights reserved.

  14. Engineered nanomaterials: toward effective safety management in research laboratories.

    Science.gov (United States)

    Groso, Amela; Petri-Fink, Alke; Rothen-Rutishauser, Barbara; Hofmann, Heinrich; Meyer, Thierry

    2016-03-15

    It is still unknown which types of nanomaterials and associated doses represent an actual danger to humans and environment. Meanwhile, there is consensus on applying the precautionary principle to these novel materials until more information is available. To deal with the rapid evolution of research, including the fast turnover of collaborators, a user-friendly and easy-to-apply risk assessment tool offering adequate preventive and protective measures has to be provided. Based on new information concerning the hazards of engineered nanomaterials, we improved a previously developed risk assessment tool by following a simple scheme to gain in efficiency. In the first step, using a logical decision tree, one of the three hazard levels, from H1 to H3, is assigned to the nanomaterial. Using a combination of decision trees and matrices, the second step links the hazard with the emission and exposure potential to assign one of the three nanorisk levels (Nano 3 highest risk; Nano 1 lowest risk) to the activity. These operations are repeated at each process step, leading to the laboratory classification. The third step provides detailed preventive and protective measures for the determined level of nanorisk. We developed an adapted simple and intuitive method for nanomaterial risk management in research laboratories. It allows classifying the nanoactivities into three levels, additionally proposing concrete preventive and protective measures and associated actions. This method is a valuable tool for all the participants in nanomaterial safety. The users experience an essential learning opportunity and increase their safety awareness. Laboratory managers have a reliable tool to obtain an overview of the operations involving nanomaterials in their laboratories; this is essential, as they are responsible for the employee safety, but are sometimes unaware of the works performed. Bringing this risk to a three-band scale (like other types of risks such as biological, radiation

  15. Report on the specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors organized by the international atomic energy agency at Oarai Engineering Centre of power reactor and nuclear development corporation, Japan, November 5-7, 1991

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1992-01-01

    As recommended by the International Working Group on Fast Reactors (IWGFR), the International Atomic Energy Agency organized a specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors. Specialists from all member countries of IWGFR-China, France, Germany, India, Italy, Japan, Russia, the United Kingdom, and the United States-participated in the meeting and made presentations as listed in Table 1. The Commission of European Communities also sent representatives to the meeting. Table 2 contains a list of participants. The meeting consisted of five sessions: (1) an overview, (2) safety characteristics of decay heal removal systems, (3) safely characteristics of reactor protection systems and reactor shutdown systems, (4) safely characteristics of reactor cores, and (5) general discussions antiformulation of recommendations

  16. 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

  17. Perspective on Secure Development Activities and Features of Safety I and C Systems

    International Nuclear Information System (INIS)

    Kang, Youngdoo; Yu, Yeong Jin; Kim, Hyungtae; Kwon, Yong il; Park, Yeunsoo; Choo, Jaeyul; Son, Jun Young; Jeong, Choong Heui

    2015-01-01

    The Enforcement Decree of the Act on Physical Protection and Radiological Emergency (ED-APPRE) was revised December 2013 to include security requirements on computer systems at nuclear facilities to protect those systems against malicious cyber-attacks. It means Cyber-Security-related measures, controls and activities of safety I and C systems against cyber-attacks shall meet the requirements of ED-APPRE. Still regulation upon inadvertent access or non-malicious modifications to the safety I and C systems is covered under the Nuclear Safety Act. The objective of this paper is to propose KINS' regulatory perspective on secure development and features against non-malicious access or modification of safety I and C systems. Secure development activities and features aim to prevent inadvertent and non-malicious access, and to prevent unwanted action from personnel or connected systems for ensuring reliable operation of safety I and C systems. Secure development activities of safety I and C systems are life cycle activities to ensure unwanted, unneeded and undocumented code is not incorporated into the systems. Secure features shall be developed, verified and qualified throughout the development life cycle

  18. Perspective on Secure Development Activities and Features of Safety I and C Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Youngdoo; Yu, Yeong Jin; Kim, Hyungtae; Kwon, Yong il; Park, Yeunsoo; Choo, Jaeyul; Son, Jun Young; Jeong, Choong Heui [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    The Enforcement Decree of the Act on Physical Protection and Radiological Emergency (ED-APPRE) was revised December 2013 to include security requirements on computer systems at nuclear facilities to protect those systems against malicious cyber-attacks. It means Cyber-Security-related measures, controls and activities of safety I and C systems against cyber-attacks shall meet the requirements of ED-APPRE. Still regulation upon inadvertent access or non-malicious modifications to the safety I and C systems is covered under the Nuclear Safety Act. The objective of this paper is to propose KINS' regulatory perspective on secure development and features against non-malicious access or modification of safety I and C systems. Secure development activities and features aim to prevent inadvertent and non-malicious access, and to prevent unwanted action from personnel or connected systems for ensuring reliable operation of safety I and C systems. Secure development activities of safety I and C systems are life cycle activities to ensure unwanted, unneeded and undocumented code is not incorporated into the systems. Secure features shall be developed, verified and qualified throughout the development life cycle.

  19. Identification of informative features for predicting proinflammatory potentials of engine exhausts.

    Science.gov (United States)

    Wang, Chia-Chi; Lin, Ying-Chi; Lin, Yuan-Chung; Jhang, Syu-Ruei; Tung, Chun-Wei

    2017-08-18

    The immunotoxicity of engine exhausts is of high concern to human health due to the increasing prevalence of immune-related diseases. However, the evaluation of immunotoxicity of engine exhausts is currently based on expensive and time-consuming experiments. It is desirable to develop efficient methods for immunotoxicity assessment. To accelerate the development of safe alternative fuels, this study proposed a computational method for identifying informative features for predicting proinflammatory potentials of engine exhausts. A principal component regression (PCR) algorithm was applied to develop prediction models. The informative features were identified by a sequential backward feature elimination (SBFE) algorithm. A total of 19 informative chemical and biological features were successfully identified by SBFE algorithm. The informative features were utilized to develop a computational method named FS-CBM for predicting proinflammatory potentials of engine exhausts. FS-CBM model achieved a high performance with correlation coefficient values of 0.997 and 0.943 obtained from training and independent test sets, respectively. The FS-CBM model was developed for predicting proinflammatory potentials of engine exhausts with a large improvement on prediction performance compared with our previous CBM model. The proposed method could be further applied to construct models for bioactivities of mixtures.

  20. Safety Analysis for Key Design Features of KALIMER-600 Design Concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

    2007-02-15

    This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

  1. Integrating Safety and Mission Assurance into Systems Engineering Modeling Practices

    Science.gov (United States)

    Beckman, Sean; Darpel, Scott

    2015-01-01

    During the early development of products, flight, or experimental hardware, emphasis is often given to the identification of technical requirements, utilizing such tools as use case and activity diagrams. Designers and project teams focus on understanding physical and performance demands and challenges. It is typically only later, during the evaluation of preliminary designs that a first pass, if performed, is made to determine the process, safety, and mission quality assurance requirements. Evaluation early in the life cycle, though, can yield requirements that force a fundamental change in design. This paper discusses an alternate paradigm for using the concepts of use case or activity diagrams to identify safety hazard and mission quality assurance risks and concerns using the same systems engineering modeling tools being used to identify technical requirements. It contains two examples of how this process might be used in the development of a space flight experiment, and the design of a Human Powered Pizza Delivery Vehicle, along with the potential benefits to decrease development time, and provide stronger budget estimates.

  2. Krypton-85 hydrofracture engineering feasibility and safety evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Peretz, F.J.; Muller, M.E.; Pan, P.Y.

    1981-07-01

    Engineering studies have been made to determine the hazards associated with the disposal of /sup 85/Kr using the hydrofracture process. To assess the hazards, an effort has been made to identify the equipment required to entrain and dissolve the noble gas into the grout stream at hydrofracture pressure (up to 350 bar). Off-the-shelf or slightly modified equipment has been identified for safe and effective compression and gas-grout mixing. Each monthly injection disposes of 1.6 x 10/sup 6/ Ci of /sup 85/Kr. By connecting only one gas cylinder to the injection system at a time, the maximum amount of krypton likely to be released as a result of equipment failure is limited to 128,000 Ci. An evaluation by Los Alamos Technical Associates shows that releasing this amount of gas in less than one hour under worst-case meteorological conditions through a 30-m stack would result in a whole-body dose of 170 millirem at a distance of 1 km from the facility. A krypton collection and recovery system can further reduce this dose to 17 millirem; increasing the distance to the site boundary to 3 km can also reduce the dose by a factor of ten. Lung and skin dose estimates are 1.6 and 120 times the whole-body dose, respectively. These are all worst-case values; releases under more typical conditions would result in a significantly lower dose. No insurmountable safety or engineering problems have been identified.

  3. Krypton-85 hydrofracture engineering feasibility and safety evaluation

    International Nuclear Information System (INIS)

    Peretz, F.J.; Muller, M.E.; Pan, P.Y.

    1981-07-01

    Engineering studies have been made to determine the hazards associated with the disposal of 85 Kr using the hydrofracture process. To assess the hazards, an effort has been made to identify the equipment required to entrain and dissolve the noble gas into the grout stream at hydrofracture pressure (up to 350 bar). Off-the-shelf or slightly modified equipment has been identified for safe and effective compression and gas-grout mixing. Each monthly injection disposes of 1.6 x 10 6 Ci of 85 Kr. By connecting only one gas cylinder to the injection system at a time, the maximum amount of krypton likely to be released as a result of equipment failure is limited to 128,000 Ci. An evaluation by Los Alamos Technical Associates shows that releasing this amount of gas in less than one hour under worst-case meteorological conditions through a 30-m stack would result in a whole-body dose of 170 millirem at a distance of 1 km from the facility. A krypton collection and recovery system can further reduce this dose to 17 millirem; increasing the distance to the site boundary to 3 km can also reduce the dose by a factor of ten. Lung and skin dose estimates are 1.6 and 120 times the whole-body dose, respectively. These are all worst-case values; releases under more typical conditions would result in a significantly lower dose. No insurmountable safety or engineering problems have been identified

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Ualikhan Zhiyenbayev; Chung, Dae Wook

    2015-01-01

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

  6. Safety outcomes for engineering asset management organizations: Old problem with new solutions?

    International Nuclear Information System (INIS)

    Novak, Jeremy; Farr-Wharton, Ben; Brunetto, Yvonne; Shacklock, Kate; Brown, Kerry

    2017-01-01

    The issue of safety and longevity of engineering assets is of increasing importance because of their impact when disasters happen. This paper addresses a literature gap by examining the role of workplace relationships in employees' safety behaviour, and builds on the Resilience Engineering (RE) framework by examining some organisational culture factors affecting how employees behave. A Social Exchange framework is used to examine the impact of supervisor-employee relationships, employee commitment to safety practices, and the type of maintenance culture upon employees’ commitment to safety and safety outcomes. Survey data from 284 technical and engineering employees in engineering asset management organisations within Australia were analyzed using Structural Equation Modelling (SEM). Effective employee relationships with management and a proactive maintenance culture were associated with employee commitment to safety culture and safety outcomes. The findings provide empirical support for embedding an effective organisational culture focused on a proactive maintenance approach, along with ensuring employees are committed to safety processes, to ensure safety outcomes and also asset longevity. One study contribution is that good safety outcomes do not develop in a vacuum; instead they are built on effective workplace relationships. Therefore, SET helps to explain the forming of effective safety culture. - Highlights: • Effective workplace relationships with management positively affect organisational safety outcomes. • Supported maintenance cultures positively affect organisational safety outcomes. • Asset longevity requires strong focus on maintenance and safety embedded in the work cultures and everyday practices of employees.

  7. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of); Choi, Young Jae; Kim, Yong Jin [KAIST, Daejeon (Korea, Republic of); Park, Hyo Chan; Park, Youn Won [BEES, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now.

  8. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    International Nuclear Information System (INIS)

    Song, Sub Lee; Choi, Young Jae; Kim, Yong Jin; Park, Hyo Chan; Park, Youn Won

    2016-01-01

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now

  9. Development of quantitative goals for inherent safety feature design and licensing

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.

    1987-01-01

    There is now considerable interest in the development of advanced fast reactors whose major focus is inherent safety. The achievement of inherent safety can be viewed from several aspects. In the Integral Fast Reactor Concept the approach is to utilize the intrinsic characteristics of pool-type liquid metal fast breeder reactors (LMFBRs) and the properties of metal fuels to integrate a high degree of inherent safety into the design. The PRISM and SAFR concepts focus on other inherent safety features. The reactors discussed above represent a radical departure from existing LWR designs as well as previous LMFBR designs (e.g., CRBRP) which are based, for the most part, on the General Design Criteria found in 10CFR50 Appendix. In view of these parallel developments (advanced reactors exploiting inherent safety and the use of quantitative goals to augment licensing), there appears to be a need to perform research on the development of methods for designing, assessing, and licensing inherent safety features in advanced reactors. The objectives of such research are outlined

  10. Quantitative Image Feature Engine (QIFE): an Open-Source, Modular Engine for 3D Quantitative Feature Extraction from Volumetric Medical Images.

    Science.gov (United States)

    Echegaray, Sebastian; Bakr, Shaimaa; Rubin, Daniel L; Napel, Sandy

    2017-10-06

    The aim of this study was to develop an open-source, modular, locally run or server-based system for 3D radiomics feature computation that can be used on any computer system and included in existing workflows for understanding associations and building predictive models between image features and clinical data, such as survival. The QIFE exploits various levels of parallelization for use on multiprocessor systems. It consists of a managing framework and four stages: input, pre-processing, feature computation, and output. Each stage contains one or more swappable components, allowing run-time customization. We benchmarked the engine using various levels of parallelization on a cohort of CT scans presenting 108 lung tumors. Two versions of the QIFE have been released: (1) the open-source MATLAB code posted to Github, (2) a compiled version loaded in a Docker container, posted to DockerHub, which can be easily deployed on any computer. The QIFE processed 108 objects (tumors) in 2:12 (h/mm) using 1 core, and 1:04 (h/mm) hours using four cores with object-level parallelization. We developed the Quantitative Image Feature Engine (QIFE), an open-source feature-extraction framework that focuses on modularity, standards, parallelism, provenance, and integration. Researchers can easily integrate it with their existing segmentation and imaging workflows by creating input and output components that implement their existing interfaces. Computational efficiency can be improved by parallelizing execution at the cost of memory usage. Different parallelization levels provide different trade-offs, and the optimal setting will depend on the size and composition of the dataset to be processed.

  11. Evaluation of features to support safety and quality in general practice clinical software

    Science.gov (United States)

    2011-01-01

    Background Electronic prescribing is now the norm in many countries. We wished to find out if clinical software systems used by general practitioners in Australia include features (functional capabilities and other characteristics) that facilitate improved patient safety and care, with a focus on quality use of medicines. Methods Seven clinical software systems used in general practice were evaluated. Fifty software features that were previously rated as likely to have a high impact on safety and/or quality of care in general practice were tested and are reported here. Results The range of results for the implementation of 50 features across the 7 clinical software systems was as follows: 17-31 features (34-62%) were fully implemented, 9-13 (18-26%) partially implemented, and 9-20 (18-40%) not implemented. Key findings included: Access to evidence based drug and therapeutic information was limited. Decision support for prescribing was available but varied markedly between systems. During prescribing there was potential for medicine mis-selection in some systems, and linking a medicine with its indication was optional. The definition of 'current medicines' versus 'past medicines' was not always clear. There were limited resources for patients, and some medicines lists for patients were suboptimal. Results were provided to the software vendors, who were keen to improve their systems. Conclusions The clinical systems tested lack some of the features expected to support patient safety and quality of care. Standards and certification for clinical software would ensure that safety features are present and that there is a minimum level of clinical functionality that clinicians could expect to find in any system.

  12. ESRS guidelines for software safety reviews. Reference document for the organization and conduct of Engineering Safety Review Services (ESRS) on software important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    The IAEA provides safety review services to assist Member States in the application of safety standards and, in particular, to evaluate and facilitate improvements in nuclear power plant safety performance. Complementary to the Operational Safety Review Team (OSART) and the International Regulatory Review Team (IRRT) services are the Engineering Safety Review Services (ESRS), which include reviews of siting, external events and structural safety, design safety, fire safety, ageing management and software safety. Software is of increasing importance to safety in nuclear power plants as the use of computer based equipment and systems, controlled by software, is increasing in new and older plants. Computer based devices are used in both safety related applications (such as process control and monitoring) and safety critical applications (such as reactor protection). Their dependability can only be ensured if a systematic, fully documented and reviewable engineering process is used. The ESRS on software safety are designed to assist a nuclear power plant or a regulatory body of a Member State in the review of documentation relating to the development, application and safety assessment of software embedded in computer based systems important to safety in nuclear power plants. The software safety reviews can be tailored to the specific needs of the requesting organization. Examples of such reviews are: project planning reviews, reviews of specific issues and reviews prior final acceptance. This report gives information on the possible scope of ESRS software safety reviews and guidance on the organization and conduct of the reviews. It is aimed at Member States considering these reviews and IAEA staff and external experts performing the reviews. The ESRS software safety reviews evaluate the degree to which software documents show that the development process and the final product conform to international standards, guidelines and current practices. Recommendations are

  13. Innovative safety features of VVER for ensuring high degree of autonomy during beyond design basis accidents

    International Nuclear Information System (INIS)

    Kumar, Abhay; Mohan, Joe; Kumar, Devesh; Chaudhry, S.M.; Rao, Srinivasa; Gupta, S.K.

    2010-01-01

    The effectiveness of Passive Heat Removal System (PHRS) in during a station black-out (SBO) accident was assessed by using SCDAP/Relap5. The analysis gave evidence that (i) the Passive Heat Removal System (PHRS) is capable of limiting the consequences of station black out (SBO) and acts as an effective engineered safety system, and (ii) the PHRS intervention prevents core degradation and excessive core heat-up. (P.A.)

  14. System safety engineering in the development of advanced surface transportation vehicles

    Science.gov (United States)

    Arnzen, H. E.

    1971-01-01

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

  15. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs

  16. Criticality safety engineering at the Savannah River Site - the 1990s

    International Nuclear Information System (INIS)

    Chandler, J.R.; Apperson, C.E. Jr.

    1996-01-01

    The privatization and downsizing effort that is ongoing within the U.S. Department of Energy (DOE) is requiring a change in the management of criticality safety engineering resources at the Savannah River Site (SRS). Downsizing affects the number of criticality engineers employed by the prime contractor, Westinghouse Savannah River Company (WSRC), and privatization affects the manner in which business is conducted. In the past, criticality engineers at the SRS have been part of the engineering organizations that support each facility handling fissile material. This practice led to different criticality safety engineering organizations dedicated to fuel fabrication activities, reactor loading and unloading activities, separation and waste management operations, and research and development

  17. Collection of methods for reliability and safety engineering

    International Nuclear Information System (INIS)

    Fussell, J.B.; Rasmuson, D.M.; Wilson, J.R.; Burdick, G.R.; Zipperer, J.C.

    1976-04-01

    The document presented contains five reports each describing a method of reliability and safety engineering. Report I provides a conceptual framework for the study of component malfunctions during system evaluations. Report II provides methods for locating groups of critical component failures such that all the component failures in a given group can be caused to occur by the occurrence of a single separate event. These groups of component failures are called common cause candidates. Report III provides a method for acquiring and storing system-independent component failure logic information. The information stored is influenced by the concepts presented in Report I and also includes information useful in locating common cause candidates. Report IV puts forth methods for analyzing situations that involve systems which change character in a predetermined time sequence. These phased missions techniques are applicable to the hypothetical ''accident chains'' frequently analyzed for nuclear power plants. Report V presents a unified approach to cause-consequence analysis, a method of analysis useful during risk assessments. This approach, as developed by the Danish Atomic Energy Commission, is modified to reflect the format and symbology conventionally used for other types of analysis of nuclear reactor systems

  18. A Resilience Engineering Approach for Sustainable Safety in Green Construction

    Directory of Open Access Journals (Sweden)

    Lucio V. Rosa

    2017-12-01

    Full Text Available Sustainable construction is a complex endeavour, involving various stakeholders and resulting in situations that are incompletely described or underspecified. Traditional risk assessment methods require a detailed description of the system and safety, focusing on undesirable outcomes, losses, incidents and accidents. Developing this principle, this research describes a new way to deal with risk assessment in the green construction industry using a resilience engineering method based on the functional resonanceanalysis method and analytic hierarchy process methodologies. The functional resonance analysis method defines a systemic framework to model complex systems based on combinations of function variabilities during normal work. Therefore, to quantify the outcomes for risk assessment, this method was used together with the analytic hierarchy process in a case study during the modernisation work on the Maracanã stadium in Rio de Janeiro. The results of this case study demonstrate that the combined utilisation of the functional resonance analysis method and analytic hierarchy process can be utilised to recognise situations where developments could potentially be without control, which enables this to be used as a basis for performing indicators or a monitoring system. Furthermore, this combined technique can be used to assess and quantify the performance variabilities that may lead to occupational or environmental accidents, and provide new recommendations about how work processes should function, minimising production losses, incidents and accidents.

  19. A new design concept for offshore nuclear power plants with enhanced safety features

    International Nuclear Information System (INIS)

    Lee, Kihwan; Lee, Kang-Heon; Lee, Jeong Ik; Jeong, Yong Hoon; Lee, Phill-Seung

    2013-01-01

    Highlights: ► A new design concept for offshore nuclear power plants is proposed. ► The total general arrangement for the concept is suggested. ► A new emergency passive containment cooling system (EPCCS) is proposed. ► A new emergency passive reactor-vessel cooling system (EPRVCS) is proposed. ► Safety features against earthquakes, tsunamis, and storms are discussed. - Abstract: In this paper, we present a new concept for offshore nuclear power plants (ONPP) with enhanced safety features. The design concept of a nuclear power plant (NPP) mounted on gravity-based structures (GBSs), which are widely used offshore structures, is proposed first. To demonstrate the feasibility of the concept, a large-scale land-based nuclear power plant model APR1400, which is the most recent NPP model in the Republic of Korea, is mounted on a GBS while minimizing modification to the original features of APR1400. A new total general arrangement (GA) and basic design principles are proposed and can be directly applied to any existing land based large scale NPPs. The proposed concept will enhance the safety of a NPP due to several aspects. A new emergency passive containment cooling system (EPCCS) and emergency passive reactor-vessel cooling system (EPRVCS) are proposed; their features of using seawater as coolant and safety features against earthquakes, Tsunamis, storms, and marine collisions are also described. We believe that the proposed offshore nuclear power plant is more robust than conventional land-based nuclear power plants and it has strong potential to provide great opportunities in nuclear power industries by decoupling the site of construction and that of installation.

  20. The LAILAPS search engine: a feature model for relevance ranking in life science databases.

    Science.gov (United States)

    Lange, Matthias; Spies, Karl; Colmsee, Christian; Flemming, Steffen; Klapperstück, Matthias; Scholz, Uwe

    2010-03-25

    Efficient and effective information retrieval in life sciences is one of the most pressing challenge in bioinformatics. The incredible growth of life science databases to a vast network of interconnected information systems is to the same extent a big challenge and a great chance for life science research. The knowledge found in the Web, in particular in life-science databases, are a valuable major resource. In order to bring it to the scientist desktop, it is essential to have well performing search engines. Thereby, not the response time nor the number of results is important. The most crucial factor for millions of query results is the relevance ranking. In this paper, we present a feature model for relevance ranking in life science databases and its implementation in the LAILAPS search engine. Motivated by the observation of user behavior during their inspection of search engine result, we condensed a set of 9 relevance discriminating features. These features are intuitively used by scientists, who briefly screen database entries for potential relevance. The features are both sufficient to estimate the potential relevance, and efficiently quantifiable. The derivation of a relevance prediction function that computes the relevance from this features constitutes a regression problem. To solve this problem, we used artificial neural networks that have been trained with a reference set of relevant database entries for 19 protein queries. Supporting a flexible text index and a simple data import format, this concepts are implemented in the LAILAPS search engine. It can easily be used both as search engine for comprehensive integrated life science databases and for small in-house project databases. LAILAPS is publicly available for SWISSPROT data at http://lailaps.ipk-gatersleben.de.

  1. Safety analysis for key design features of KALIMER-600 design concept

    International Nuclear Information System (INIS)

    Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S.

    2005-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed

  2. Nuclear power plants near consumers from a safety-engineering point of view

    International Nuclear Information System (INIS)

    Kroeger, W.

    1986-11-01

    Special safety requirements must be met by a nuclear power station near the consumer. These requirements may not be formulated in a purely probabilistic way because of the methodological deficiencies identified. The existing protection concept is rather extended so as to include the requirement of engineered safeguards in order to limit the damage in case of a worst reactor accident. The suggested individual dose limit together with the calculation rules should ensure that the consequences of a worst accident are essentially limited to the plant and that no emergency protection measures and countermeasures need to be considered either in the short term or in the longer term to prevent health damage. The resulting features of a reactor near the consumer aim at better inherent safety characteristics, which is shown to be possible by reasonable technical means and which seems to have already been realized to a large extent in plants of small and perhaps also medium power already conceived. The way of thinking behind this suggestion is applicable to other sectors of industrial technology. Furthermore, it might serve as a basis in the discussion about general advanced safety criteria, which has been stimulated due to 'Chernobyl'. (orig./HP) [de

  3. LOFT integral test system final safety analysis report

    International Nuclear Information System (INIS)

    1974-03-01

    Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)

  4. Design and safety features of commercial nuclear power plants in Japan, 1976 edition

    International Nuclear Information System (INIS)

    Izumi, Fumio; Harayama, Yasuo

    1976-10-01

    The December 1975 edition (JAERI-M 5959) contained design particulars and safety features of 20 commercial nuclear power plants in Japan as of December 1974. Subsequently new plants have been put into operation and some plants under construction have undergone design modifications. The present edition presents similar data of the commercial nuclear power plants in Japan up to June 1976, compiled by computer processing. (auth.)

  5. Cost reduction and safety design features of ABWR-II. Annex 5

    International Nuclear Information System (INIS)

    Koh, F.; Moriya, K.; Anegawa, T.

    2002-01-01

    The ABWR-II, which is aimed to be the next generation reactor following the latest BWR: Advanced Boiling Reactor (ABWR), is now under development jointly by the Japanese BWR utilities, General Electric Company, Hitachi Limited, and Toshiba Corporation. The key objectives of ABWR-II development include improvement in economics and further sophistication in safety for commercialization in the late 2010's and after. This paper summarizes the current status of ABWR-II development focusing on economics and safety. Plant power rating, fuel size, CRD rationalization and outage period are discussed from a cost reduction perspective. In terms of safety, the features such as diversification in emergency power sources and passive system application against severe accidents are being introduced. (author)

  6. Main design and safety features of a 200MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zheng, Wenxiang; Gao, Zuying; Wang, Dazhong

    1992-01-01

    Inept has been in charge of the development of a nuclear heating reactor since 1980s, which is one of the national key R and D Programs in China. A 5MWt experimental NCR was completed at Inept in 1989 and has operated successfully for space heating since then. In order to realize the commercialization of the NCR, it has been decided to construct a 200MW demonstration NCR in 1993. A number of advanced features, including natural circulation, integrated arrangement, self-pressurized performance, dual vessel structure, hydraulic control rod drive and passive safety systems, have been incorporated into the NCR-200 to achieve its safety goal and economic viability. This makes the NCR safe, simple, reliable, easy-constructed and maintained. At present, the design work of the NCR-200 have shown that its safety characteristics are excellent. The NCR could play an important role in resolving future energy and environmental problems in China. The paper will mainly cover the key design considerations, main technical features and safety analysis results of the NCR-200

  7. Safety Concepts in Structural Glass Engineering : Towards an Integrated Approach

    NARCIS (Netherlands)

    Bos, F.P.

    2009-01-01

    This dissertation proposes the Integrated Approach to Structural Glass Safety, based on four clearly defined element safety properties, damage sensitivity, relative resistance, redundancy, and fracture mode. The Element Safety Diagram (ESD) is introduced to provide an easy-to-read graphical

  8. Safety review for human factors engineering and control rooms of nuclear power plants

    International Nuclear Information System (INIS)

    Yang Mengzhuo

    1998-01-01

    Safety review for human factors engineering and control rooms of nuclear power plants (NPP) is in a forward position of science and technology, which began at American TMI severe accident and had been implemented in China. The importance and the significance of the safety review are expounded, the requirements of its scope and profundity are explained in detail. In addition, the situation of the technical document system for nuclear safety regulation on human factors engineering and control rooms of NPP in China is introduced briefly, on which the safety review is based

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-06-01

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

  10. Problems of Formation of Diagnostic Features in the Diagnosis of Aircraft Engines

    Directory of Open Access Journals (Sweden)

    Pronyakin V.I.

    2017-01-01

    Full Text Available The article is devoted to the evaluation of current technical condition of aircraft engines. Deals with the choice of the detection method of diagnostic features required for degradation assessment, emergency protection and detection of incipient defects on the example of cyclic machines and mechanisms. For the formation of diagnostic features in the diagnosis of aircraft engines use different physical effects (vibration, shock, heat radiation, electrodynamic and thermal processes, wear debris in oil, etc.. Classification of defects and requirements for the development of diagnostics systems is formed based on them. The article describes the requirements for diagnostic signs. The article provides a promising phase method that allows obtaining stable diagnostic characters in exploitation. The result of applying the method is shown. Diagnostic signs are formed. In mathematical modeling it was used the traditional theory of the description of rotary mechanisms. The data obtained are compared with experimental data.

  11. The feature of high flux engineering test reactor and its role in nuclear power development

    International Nuclear Information System (INIS)

    Lu Guangquan

    1987-01-01

    The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

  12. Evaluation of special safety features of the SNR-300 in view of the Chernobyl accident

    International Nuclear Information System (INIS)

    Vossebrecker, H.

    1987-03-01

    A comparison of those characteristics, which decisively influenced the accident in the RMBK-1000 reactor, with the safety features of SNR-300 has been performed. The conclusions of this comparison are presented in the present report. The SNR-300 is characterized by a stable reactivity behaviour and good controllability, whereas RBMK-1000 has an instable behaviour and complex spatial dependencies in the core. Among other points, design deficiencies in the protection and emergency shutdown systems were responsible for the Chernobyl accident. The protection and scram systems of the SNR-300 are unquestionably superior to those of the RBMK-1000 with regard to redundancy, diversity, degree of automation, separation of operational and safety-relevant tasks, protection against inadmissible interventions, effectiveness and safety reserves. Therefore, excursion accidents can be classified as hypothetical for SNR-300. Due to elementary physical properties, possible energy releases during hypothetical excursions are substantially lower for SNR-300 and would be controlled by the design of the primary system and containment systems. No damage limiting measures are provided in the RBMK-100 for excursion accidents. Finally, exothermal processes augmented the consequences of the accident in the RBMK-1000 and the long-lasting graphite fire intensified the release of radioactivity. In the SNR-300, however, inertisation of the containment, the steel plate lining and the floor troughs ensure that activity enclosure inside the containment after leakage or hypothetical excursion accident is not endangered by exothermal reactions. Further safety aspects are presented in the report, which can be linked with the accident in Chernobyl. In summary, it is obvious that the disadvantageous physical and technical features of the RBMK-1000 do either not exist in the SNR-300 or are covered by the safety design

  13. Effects of exhaust gas recirculation in diesel engines featuring late PCCI type combustion strategies

    International Nuclear Information System (INIS)

    D’Ambrosio, S.; Ferrari, A.

    2015-01-01

    Highlights: • The effects that a high EGR rate can have on PCCI type combustion strategies have been analyzed. • The dependence of engine emissions and combustion noise on EGR has been addressed. • The time histories of the main in-cylinder variables have been plotted for different EGR rates. - Abstract: The influence of exhaust gas recirculation (EGR) has been analyzed considering experimental results obtained from a Euro 5 diesel engine calibrated with an optimized pilot-main double injection strategy. The engine features a late premixed charge compression ignition (PCCI) type combustion mode. Different steady-state key-points that are representative of the engine application in a passenger car over the New European Driving Cycle (NEDC) have been studied. The engine was fully instrumented to obtain a complete overview of the most important variables. The pressure time history in the combustion chamber has been measured to perform calculations with single and three-zone combustion diagnostic models. These models allow the in-cylinder emissions and the temperature of the burned and unburned zones to be evaluated as functions of the crankshaft angle. The EGR mass fraction was experimentally varied within the 0–50% range. The results of the investigation have shown the influence that high EGR rates can have on intake and exhaust temperatures, in-cylinder pressure and heat release rate time histories, engine-out emissions (CO, HC, NO_x, soot), brake specific fuel consumption and combustion noise for a PCCI type combustion strategy. The outputs of the diagnostic models have been used to conduct a detailed analysis of the cause-and-effect relationships between the EGR rate variations and the engine performance. Finally, the effect of the EGR on the cycle-to-cycle variability of the engine torque has been experimentally investigated.

  14. A Co-modeling Method Based on Component Features for Mechatronic Devices in Aero-engines

    Science.gov (United States)

    Wang, Bin; Zhao, Haocen; Ye, Zhifeng

    2017-08-01

    Data-fused and user-friendly design of aero-engine accessories is required because of their structural complexity and stringent reliability. This paper gives an overview of a typical aero-engine control system and the development process of key mechatronic devices used. Several essential aspects of modeling and simulation in the process are investigated. Considering the limitations of a single theoretic model, feature-based co-modeling methodology is suggested to satisfy the design requirements and compensate for diversity of component sub-models for these devices. As an example, a stepper motor controlled Fuel Metering Unit (FMU) is modeled in view of the component physical features using two different software tools. An interface is suggested to integrate the single discipline models into the synthesized one. Performance simulation of this device using the co-model and parameter optimization for its key components are discussed. Comparison between delivery testing and the simulation shows that the co-model for the FMU has a high accuracy and the absolute superiority over a single model. Together with its compatible interface with the engine mathematical model, the feature-based co-modeling methodology is proven to be an effective technical measure in the development process of the device.

  15. 67. The safety engineering at driving of destroyed hearth and repair of bath fettling during operation

    International Nuclear Information System (INIS)

    Ivanov, A.V.

    1993-01-01

    The safety engineering at driving of destroyed hearth and repair of bath fettling during operation was considered. All operational conditions at driving of destroyed hearth and repair of bath fettling during operation were studied.

  16. Environmental, Health, and Safety Research Needs for Engineered Nanoscale Materials

    National Research Council Canada - National Science Library

    Alderson, Norris; Alexander, Catherine; Merzbacher, Celia; Chernicoff, William; Middendorf, Paul; Beck, Nancy; Chow, Flora; Poster, Dianne; Danello, Mary Ann; Barrera, Enriqueta

    2006-01-01

    ...) research and information needs related to understanding and management of potential risks of engineered nanoscale materials that may be used, for example, in commercial or consumer products, medical...

  17. 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

  18. Incorporation of Safety into Design Process : A Systems Engineering Perspective

    NARCIS (Netherlands)

    Rajabalinejad, M.

    2018-01-01

    This paper suggests integrating the best safety practices with the design process. This integration enriches the exploration experience for designers and adds extra values and competitor advantages for customers. The paper introduces the safety cube for combining common blocks for design, hazard

  19. Interaction between systems and software engineering in safety-critical systems

    International Nuclear Information System (INIS)

    Knight, J.

    1994-01-01

    There are three areas of concern: when is software to be considered safe; what, exactly, is the role of the software engineer; and how do systems, or sometimes applications, engineers and software engineers interact with each other. The author presents his perspective on these questions which he feels differ from those of many in the field. He argues for a clear definition of safety in the software arena, so the engineer knows what he is engineering toward. Software must be viewed as part of the entire system, since it does not function on its own, or isolation. He argues for the establishment of clear specifications in this area

  20. Report to NASA Committee on Aircraft Operating Problems Relative to Aviation Safety Engineering and Research Activities

    Science.gov (United States)

    1963-01-01

    The following report highlights some of the work accomplished by the Aviation Safety Engineering and Research Division of the Flight Safety Foundations since the last report to the NASA Committee on Aircraft Operating Problems on 22 May 1963. The information presented is in summary form. Additional details may be provided upon request of the reports themselves may be obtained from AvSER.

  1. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

  2. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    International Nuclear Information System (INIS)

    2012-12-01

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

  3. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

  4. Safety margin improvement by adopting the feature of interleaving in 700 MWe PHWR

    International Nuclear Information System (INIS)

    Kumar, Nrependra; Yadav, S.K.; Khan, T.A.; Dixit, A.; Singhal, Mukesh; Nair, Suma R.

    2015-01-01

    Indian Pressurised Heavy Water Reactors (IPHWRs) of 700 MWe are under construction at Kakrapar Atomic Power Project -3,4 and Rajasthan Atomic Power Project-7,8. These units have enhanced safety features with respect to standard IPHWRs. One of the enhanced features is interleaving of feeders/channels. In interleaved feeder configuration, each header located at either end of reactor gets connected to one quarter of core channels, which are uniformly distributed. The core is divided into two loops with feeder connected in interleaved fashioned. In this paper a comparative study has been performed between the two cases: 1) The core splits in two vertical halves and each vertical half is a loop of PHT (TAPS-3 and 4 Type configuration). 2) The core is divided into two loops with feeders/ channels connected in interleaved fashioned (700 MWe Configuration). LOCA studies have been performed for 700 MWe PHWR considering interleaving of feeders configuration using in-house developed computer code ATMIKA and 3-D neutron kinetics code IQS-3D. The issue of interleaving is closely linked to an inherent reactivity characteristic of PHWR reactors (viz., positive void reactivity coefficient) which leads to a power increase following a Large LOCA. In 700 MWe PHWR with intent to improve the safety margin, adopted the feature of interleaving of feeders which causes in reduction in the magnitude of void coefficient and results in reduction of peak power during LBLOCA. The systematic LBLOCA study demonstrates that interleaved configuration of feeder/channels of two loops has higher safety margins (i.e. with respect to peak power, prompt-criticality margin, adiabatic heat deposition on the fuel pins, sheath temperature excursion and clad oxidation) with regard to the effectiveness of shutdown system. (author)

  5. Conservation of Life as a Unifying Theme for Process Safety in Chemical Engineering Education

    Science.gov (United States)

    Klein, James A.; Davis, Richard A.

    2011-01-01

    This paper explores the use of "conservation of life" as a concept and unifying theme for increasing awareness, application, and integration of process safety in chemical engineering education. Students need to think of conservation of mass, conservation of energy, and conservation of life as equally important in engineering design and analysis.…

  6. 76 FR 78 - Federal Motor Vehicle Safety Standard; Engine Control Module Speed Limiter Device

    Science.gov (United States)

    2011-01-03

    ... [Docket No. NHTSA-2007-26851] Federal Motor Vehicle Safety Standard; Engine Control Module Speed Limiter... occupants. IIHS stated that on-board electronic engine control modules (ECM) will maintain the desired speed... be equipped with an electronic control module (ECM) that is capable of limiting the maximum speed of...

  7. 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

  8. Application of system safety engineering techniques for hazard prevention at the Superconducting Super Collider

    International Nuclear Information System (INIS)

    Hendrix, B.L.

    1991-01-01

    A primary goal of the Superconducting Super Collider Laboratory (SSCL) is to establish an exemplary safety program. Achieving this goal requires leadership, planning, coordination, and technical know-how. To ensure that safety is an inherent part of the design, the Environment, Safety and Health Office employs a systems engineering discipline and process known as System Safety. The goal of System Safety - hazard prevention - is accomplished by analyzing systems to identify hazards and to evaluate design and procedural options and countermeasures to prevent, eliminate, mitigate, or control hazards and risks. Establishment of safety and human factors design criteria at the outset of the project prevents unsafe designs and safety violations, reduces risks, and helps in avoiding costly design changes later. This process requires a considerable amount of coordination with a variety of technical disciplines and safety professionals to integrate methods of hazard prevention, mitigation, and risk reduction throughout the system life-cycle

  9. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Su, S.F.; Cahalan, J.E.; Sevy, R.H.

    1985-01-01

    This paper presents the results of a systematic study of the effectiveness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). The results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs, and safety margins are quantified in sensitivity studies. All analyses were carried out using the SASSYS LMFBR systems analysis code (1)

  10. Application of software engineering to development of reactor-safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1980-11-01

    As a result of the drastically increasing cost of software and the lack of an engineering approach, the technology of Software Engineering is being developed. Software Engineering provides an answer to the increasing cost of developing and maintaining software. It has been applied extensively in the business and aerospace communities and is just now being applied to the development of scientific software and, in particular, to the development of reactor safety codes at HEDL

  11. Safety leadership in the teaching laboratories of electrical and electronic engineering departments at Taiwanese Universities.

    Science.gov (United States)

    Wu, Tsung-Chih

    2008-01-01

    Safety has always been one of the principal goals in teaching laboratories. Laboratories cannot serve their educational purpose when accidents occur. The leadership of department heads has a major impact on laboratory safety, so this study discusses the factors affecting safety leadership in teaching laboratories. This study uses a mail survey to explore the perceived safety leadership in electrical and electronic engineering departments at Taiwanese universities. An exploratory factor analysis shows that there are three main components of safety leadership, as measured on a safety leadership scale: safety controlling, safety coaching, and safety caring. The descriptive statistics also reveals that among faculty, the perception of department heads' safety leadership is in general positive. A two-way MANOVA shows that there are interaction effects on safety leadership between university size and instructor age; there are also interaction effects between presence of a safety committee and faculty gender and faculty age. It is therefore necessary to assess organizational factors when determining whether individual factors are the cause of differing perceptions among faculty members. The author also presents advice on improving safety leadership for department heads at small universities and at universities without safety committees.

  12. Integration of safety engineering into a cost optimized development program.

    Science.gov (United States)

    Ball, L. W.

    1972-01-01

    A six-segment management model is presented, each segment of which represents a major area in a new product development program. The first segment of the model covers integration of specialist engineers into 'systems requirement definition' or the system engineering documentation process. The second covers preparation of five basic types of 'development program plans.' The third segment covers integration of system requirements, scheduling, and funding of specialist engineering activities into 'work breakdown structures,' 'cost accounts,' and 'work packages.' The fourth covers 'requirement communication' by line organizations. The fifth covers 'performance measurement' based on work package data. The sixth covers 'baseline requirements achievement tracking.'

  13. Human factors and systems engineering approach to patient safety for radiotherapy.

    Science.gov (United States)

    Rivera, A Joy; Karsh, Ben-Tzion

    2008-01-01

    The traditional approach to solving patient safety problems in healthcare is to blame the last person to touch the patient. But since the publication of To Err is Human, the call has been instead to use human factors and systems engineering methods and principles to solve patient safety problems. However, an understanding of the human factors and systems engineering is lacking, and confusion remains about what it means to apply their principles. This paper provides a primer on them and their applications to patient safety.

  14. Assessment of NPP safety taking into account seismic and engineering-geological factors

    International Nuclear Information System (INIS)

    Yakovlev, E.A.

    1990-01-01

    Consideration is given to the problem of probabilistic analysis of NPP safety with account of risk of destructive effect of earthquakes and the danger of accidental geological processes (diapirism, karst etc.) under NPP operation. It is shown that account of seismic and engineering-geological (engineering-seismological) risk factors in probabilistic analysis of safety enables to perform anticipatory analysis of behaviour of principle plant objects and to improve safety of their operation by revealing the most unstable elements of geotechnical system forming the main contribution to the total NPP risk

  15. Human Factors and Systems Engineering Approach to Patient Safety for Radiotherapy

    International Nuclear Information System (INIS)

    Rivera, A. Joy; Karsh, Ben-Tzion

    2008-01-01

    The traditional approach to solving patient safety problems in healthcare is to blame the last person to touch the patient. But since the publication of To Err is Human, the call has been instead to use human factors and systems engineering methods and principles to solve patient safety problems. However, an understanding of the human factors and systems engineering is lacking, and confusion remains about what it means to apply their principles. This paper provides a primer on them and their applications to patient safety

  16. A regression-based Kansei engineering system based on form feature lines for product form design

    Directory of Open Access Journals (Sweden)

    Yan Xiong

    2016-06-01

    Full Text Available When developing new products, it is important for a designer to understand users’ perceptions and develop product form with the corresponding perceptions. In order to establish the mapping between users’ perceptions and product design features effectively, in this study, we presented a regression-based Kansei engineering system based on form feature lines for product form design. First according to the characteristics of design concept representation, product form features–product form feature lines were defined. Second, Kansei words were chosen to describe image perceptions toward product samples. Then, multiple linear regression and support vector regression were used to construct the models, respectively, that predicted users’ image perceptions. Using mobile phones as experimental samples, Kansei prediction models were established based on the front view form feature lines of the samples. From the experimental results, these two predict models were of good adaptability. But in contrast to multiple linear regression, the predict performance of support vector regression model was better, and support vector regression is more suitable for form regression prediction. The results of the case showed that the proposed method provided an effective means for designers to manipulate product features as a whole, and it can optimize Kansei model and improve practical values.

  17. A Comparison of the mechanical engineering and safety engineering student’s ICT attitudes at the Obuda University

    Directory of Open Access Journals (Sweden)

    Kiss Gabor

    2016-01-01

    Full Text Available Communication and technology are critical to education. However, using technology in education is not an easy task as communication barriers emerge. The aim of this research is to analyze the ICT attitudes from different faculties at the Obuda University that is between the mechanical engineering students and safety engineering students from the Donát Bánki Mechanical Safety Engineer Faculty. The students from these two groups will use different ICT tool at work after their graduation; the mechanical engineering students will work mostly with designer ICT tools, the safety engineering students will use security systems. It would be important to know whether instructors, when using ICT, have to follow different teaching methods and approaches in these two different groups or not. We measured the ICT attitude with a tool consisting of 23 items (Likert scaled. We worked with 361 students. The data analysis was performed with SPSS software using descriptive statistics and Mann-Whitney test. The results show both groups having the same positive ICT attitude however with one difference.

  18. Leading Edge. Volume 7, Number 3. Systems Safety Engineering

    Science.gov (United States)

    2010-01-01

    foods were not always safe to eat given the sanitary conditions of the day. In 1943, the psychologist Abraham Maslow proposed a five-level... hierarchy of basic human needs, and safety was number two on this list. System safety is a specialized and formalized extension of our in- herent drive for...factors, hazards, mishaps, and ef- fects. The following is an example of each element within the hierarchy : An exposed sharp edge in a relay cabi- net

  19. Rail Safety/Equipment Crashworthiness : Volume 3. Proposed Engineering Standards.

    Science.gov (United States)

    1978-07-01

    The document, the third of four volumes, contains recommended Engineering Standards prepared in the format of the standards published in the Code of Federal Regulations (Title 49, Transportation, Parts 200). The standards proposed provide improved oc...

  20. Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: vasconv@cdtn.br, e-mail: silvaem@cdtn.br, e-mail: aclc@cdtn.br, e-mail: reissc@cdtn.br

    2009-07-01

    Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

  1. Safety, reliability, risk management and human factors: an integrated engineering approach applied to nuclear facilities

    International Nuclear Information System (INIS)

    Vasconcelos, Vanderley de; Silva, Eliane Magalhaes Pereira da; Costa, Antonio Carlos Lopes da; Reis, Sergio Carneiro dos

    2009-01-01

    Nuclear energy has an important engineering legacy to share with the conventional industry. Much of the development of the tools related to safety, reliability, risk management, and human factors are associated with nuclear plant processes, mainly because the public concern about nuclear power generation. Despite the close association between these subjects, there are some important different approaches. The reliability engineering approach uses several techniques to minimize the component failures that cause the failure of the complex systems. These techniques include, for instance, redundancy, diversity, standby sparing, safety factors, and reliability centered maintenance. On the other hand system safety is primarily concerned with hazard management, that is, the identification, evaluation and control of hazards. Rather than just look at failure rates or engineering strengths, system safety would examine the interactions among system components. The events that cause accidents may be complex combinations of component failures, faulty maintenance, design errors, human actions, or actuation of instrumentation and control. Then, system safety deals with a broader spectrum of risk management, including: ergonomics, legal requirements, quality control, public acceptance, political considerations, and many other non-technical influences. Taking care of these subjects individually can compromise the completeness of the analysis and the measures associated with both risk reduction, and safety and reliability increasing. Analyzing together the engineering systems and controls of a nuclear facility, their management systems and operational procedures, and the human factors engineering, many benefits can be realized. This paper proposes an integration of these issues based on the application of systems theory. (author)

  2. Factors Affecting the Behavior of Engineering Students toward Safety Practices in the Machine Shop

    Directory of Open Access Journals (Sweden)

    Jessie Kristian M. Neria

    2015-08-01

    Full Text Available This study aimed to determine the factors that affect the behavior of engineering student toward safety practices in the machine shop. Descriptive type of research was utilized in the study. Results showed that most of the engineering students clearly understand the signage shown in the machine shop. Students are aware that they should not leave the machines unattended. Most of the engineering students handle and use the machine properly. The respondents have an average extent of safety practices in the machine shop which means that they are applying safety practices in their every activity in machine shop. There is strong relationship between the safety practices and the factors affecting behavior in terms of signage, reminder of teacher and rules and regulation.

  3. Expertise preservation in nuclear technology - the new master course ''nuclear safety engineering'' at the RWTH Aachen

    International Nuclear Information System (INIS)

    Backus, Sabine; Heuters, Michael

    2011-01-01

    The energy concept of the German federal Government in 2010 emphasizes the importance of nuclear energy within the energy policy. The lifetime extension of German nuclear power plants and the long-term safety of radioactive waste storage is the new challenge with respect to the expertise preservation in Germany. The owners of nuclear utilities have started to assist new research programs in the field of nuclear engineering at the German universities. RWE Power and ThyssenKrupp have signed a cooperation contract in 2007 with the RWTH Aachen. The companies bear the expenses for professorships ''nuclear fuel cycle'', ''simulation in nuclear engineering'' and ''reactor safety and engineering''. An elongation of the contract is planned. A master course ''nuclear safety engineering'' over 4 semesters covers the complete fuel cycle. The authors discuss issues concerning the information of students, experiences with the expectations of students concerning their future employment, acceptance of nuclear energy and related topics.

  4. Understanding safety and production risks in rail engineering planning and protection.

    Science.gov (United States)

    Wilson, John R; Ryan, Brendan; Schock, Alex; Ferreira, Pedro; Smith, Stuart; Pitsopoulos, Julia

    2009-07-01

    Much of the published human factors work on risk is to do with safety and within this is concerned with prediction and analysis of human error and with human reliability assessment. Less has been published on human factors contributions to understanding and managing project, business, engineering and other forms of risk and still less jointly assessing risk to do with broad issues of 'safety' and broad issues of 'production' or 'performance'. This paper contains a general commentary on human factors and assessment of risk of various kinds, in the context of the aims of ergonomics and concerns about being too risk averse. The paper then describes a specific project, in rail engineering, where the notion of a human factors case has been employed to analyse engineering functions and related human factors issues. A human factors issues register for potential system disturbances has been developed, prior to a human factors risk assessment, which jointly covers safety and production (engineering delivery) concerns. The paper concludes with a commentary on the potential relevance of a resilience engineering perspective to understanding rail engineering systems risk. Design, planning and management of complex systems will increasingly have to address the issue of making trade-offs between safety and production, and ergonomics should be central to this. The paper addresses the relevant issues and does so in an under-published domain - rail systems engineering work.

  5. Cost reduction and safety design features of new nuclear power plants in India. Annex 13

    International Nuclear Information System (INIS)

    Sharma, V.K.

    2002-01-01

    Indian Nuclear Power Programme is designed to exploit limited reserves of uranium and extensive resource of thorium. Pressurised heavy water reactors are found most suitable and form the main stay of the first stage of the programme. Thorium utilisation is achieved in the second and third stages. Today India has total installed capacity of 2720 MWe of PHWRs which are operating with high plant load factors of over 80%. Rich experience of construction and operation of over 150 reactor years is being utilised in effecting cost reduction and safety improvements. Standardisation and reduction in gestation period by preproject activities, advance procurement and work packages of engineer, procure, construct and commission are some of the techniques being adopted for cost reduction in the new projects. But the cost of safety is rising. Design basis event of double ended guillotine rupture of primary pressure boundary needs a relook based on current knowledge of material behaviour. This event appears improbable. Similarly some of the safety related systems like closed loop cooling water operating at low temperature and pressure, and low usage factors may be designed as per standard codes without invoking special nuclear requirements. The paper will address these issues and highlight the possible areas for cost reduction both in operating and safety systems. Modern construction and project management techniques are being employed. Gestation period of 5 years and cost of less than US $1400 per KWe are the present targets. In Indian environment nuclear power is found to be competitive with thermal power plants at distances of about 800 Kms from the coal mines. (author)

  6. Improvement of inherent safety features in CSR (Coupled Spectrum Reactor) for treating MA

    International Nuclear Information System (INIS)

    Aziz, F.; Kitamoto, Asashi.

    1996-01-01

    Burning and/or transmutation (B/T) of MA is proposed here using a CSR (Coupled Spectrum Reactor) concept. CSR was based on a modified conventional 1150 MWe-PWR system, and consisted of two core regions for thermal and fast neutrons, respectively. The B/T fuel used was supposed such that MA discharged from 1 GWe-LWR were mixed homogeneously in LWR fuel. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio, (V m /V f ). In order to improve its inherent safety features, several cases of CSR were studied and compared, each case used different fuel type in the inner region. The result of the calculations showed that safety features can be improved by using composite fuel of ( 235 U-Pu- 238 U) in the inner region. The equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute MA up to 808 kg/stage in a single reactor operated with a reactivity swing of 2.8 % Δk/kk'. (author)

  7. Use of the self-organizing feature map to diagnose abnormal engineering change

    Science.gov (United States)

    Lu, Ruei-Shan; Wu, Zhi-Ting; Peng, Kuo-Wei; Yu, Tai-Yi

    2015-07-01

    This study established identification manners with self-organizing feature map (SOM) to achieve the goal of monitoring Engineering Change (EC) based on historical data of a company that specializes in computers and peripherals. The product life cycle of this company is 3-6 months. The historical data were divided into three parts, each covering four months. The first part, comprising 2,343 records from January to April (the training period), comprise the Control Group. The second and third parts comprise Experimental Groups (EG) 1 and 2, respectively. For EG 1 and 2, the successful rate of recognizing information on abnormal ECs was approximately 96% and 95%, respectively. This paper shows the importance and screening procedures of abnormal engineering change for a particular company specializing in computers and peripherals.

  8. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, N.; Kawatta, N.; Niwa, H.; Kondo, S.; Maeda, K

    1996-12-31

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given.

  9. Development of an Owner Engineer's independent capability in NPP safety and licensing

    International Nuclear Information System (INIS)

    Auglaire, M.; Bayart, D.; D'Eer, A.; Polet, F.; Vanhoenacker, L.; Zhang, J.

    2002-01-01

    As Owner's Engineer to Electrabel, the Belgian utility which owns and operates the 7 NPPs in Belgium, Tractebel Energy Engineering has gained considerable experience in the field of ten-yearly safety overhauls of NPPs since 1983. It has developed a methodology leading to proposing corrective actions by means of a global and integrated approach in which safety improvement costs are optimized. Safety issues addressed during those projects encompass the writing of Probabilistic Safety Assessment studies, post-TMI recommendations implementation, the installation of autocatalytic recombiners, accident studies, protection against pressurized thermal shock, impact of flooding of internal or external origin, implementation of severe accident management guidelines, re-evaluation of the environment, verification of extreme climate conditions, updating of the Safety Analysis Reports, operation review. (author)

  10. Safety assessment of complex engineered and natural systems: radioactive waste disposal

    International Nuclear Information System (INIS)

    McNeish, J.A.; Vallikat, V.; Atkins, J.; Balady, M.A.

    1997-01-01

    Evaluation of deep, geologic disposal of nuclear waste requires the probabilistic safety assessment of a complex system from the coupling of various processes and sub-systems, parameter and model uncertainties, spatial and temporal variabilities, and the multiplicity of designs and scenarios. Both the engineered and natural system are included in the evaluation. Each system has aspects with considerable uncertainty both in important parameters and in overall conceptual models. The study represented herein provides a probabilistic safety assessment of a potential respository system for multiple engineered barrier system (EBS) design and conceptual model configurations (CRWMS M and O, 1996a) and considers the effects of uncertainty on the overall results. The assessment is based on data and process models available at the time of the study and doesnt necessarily represent the current safety evaluation. In fact, the percolation flux through the repository system is now expected to be higher than the estimate used for this study. The potential effects of higher percolation fluxes are currently under study. The safety of the system was assessed for both 10,000 and 1,000,000 years. Use of alternative conceptual models also produced major improvement in safety. For example, use of a more realistic engineered system release model produced improvement of over an order of magnitude in safety. Alternative measurement locations for the safety assessment produced substantial increases in safety, through the results are based on uncertain dilution factors in the transporting groundwater. (Author)

  11. Engineering Hematopoietic Cells for Cancer Immunotherapy: Strategies to Address Safety and Toxicity Concerns.

    Science.gov (United States)

    Resetca, Diana; Neschadim, Anton; Medin, Jeffrey A

    2016-09-01

    Advances in cancer immunotherapies utilizing engineered hematopoietic cells have recently generated significant clinical successes. Of great promise are immunotherapies based on chimeric antigen receptor-engineered T (CAR-T) cells that are targeted toward malignant cells expressing defined tumor-associated antigens. CAR-T cells harness the effector function of the adaptive arm of the immune system and redirect it against cancer cells, overcoming the major challenges of immunotherapy, such as breaking tolerance to self-antigens and beating cancer immune system-evasion mechanisms. In early clinical trials, CAR-T cell-based therapies achieved complete and durable responses in a significant proportion of patients. Despite clinical successes and given the side effect profiles of immunotherapies based on engineered cells, potential concerns with the safety and toxicity of various therapeutic modalities remain. We discuss the concerns associated with the safety and stability of the gene delivery vehicles for cell engineering and with toxicities due to off-target and on-target, off-tumor effector functions of the engineered cells. We then overview the various strategies aimed at improving the safety of and resolving toxicities associated with cell-based immunotherapies. Integrating failsafe switches based on different suicide gene therapy systems into engineered cells engenders promising strategies toward ensuring the safety of cancer immunotherapies in the clinic.

  12. Prevent recurrence of nuclear disaster (3). Agenda on nuclear safety from earthquake engineering

    International Nuclear Information System (INIS)

    Kameda, Hiroyuki; Takada, Tsuyoshi; Ebisawa, Katsumi; Nakamura, Susumu

    2012-01-01

    Based on results of activities of committee on seismic safety of nuclear power plants (NPPs) of Japan Association for Earthquake Engineering, which started activities after Chuetsu-oki earthquake and then experienced Great East Japan Earthquake, (under close collaboration with the committee of Atomic Energy Society of Japan started activities simultaneously), and taking account of further development of concept, agenda on nuclear safety were proposed from earthquake engineering. In order to prevent recurrence of nuclear disaster, individual technical issues of earthquake engineering and comprehensive issues of integration technology, multidisciplinary collaboration and establishment of technology governance based on them were of prime importance. This article described important problems to be solved; (1) technical issues and mission of seismic safety of NPPs, (2) decision making based on risk assessment - basis of technical governance, (3) framework of risk, design and regulation - framework of required technology governance, (4) technical issues of earthquake engineering for nuclear safety, (5) role of earthquake engineering in nuclear power risk communication and (6) importance of multidisciplinary collaboration. Responsibility of engineering would be attributed to establishment of technology governance, cultivation of individual technology and integration technology, and social communications. (T. Tanaka)

  13. 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

  14. Nanotechnologies, engineered nanomaterials and occupational health and safety - A review

    NARCIS (Netherlands)

    Savolainen, K.; Pylkkänen, L.; Norppa, H.; Falck, G.; Lindberg, H.; Tuomi, T.; Vippola, M.; Alenius, H.; Hämeri, K.; Koivisto, J.; Brouwer, D.; Mark, D.; Bard, D.; Berges, M.; Jankowska, E.; Posniak, M.; Farmer, P.; Singh, R.; Krombach, F.; Bihari, P.; Kasper, G.; Seipenbusch, M.

    2010-01-01

    The significance of engineered nanomaterials (ENM) and nanotechnologies grows rapidly. Nanotechnology applications may have a positive marked impact on many aspects of on human every day life, for example by providing means for the production of clean energy and pure drinking water. Hundreds of

  15. Design data and safety features of commercial nuclear power plants. Vol. IV. Dockets 50-452 through 50-503

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1975-03-01

    Design data, safety features, and site characteristics are summarized for 36 nuclear power units in 18 power stations in the United States. Six pages of data are presented for each plant consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. The volume covers reactors with dockets 50-452 through 50-503. (U.S.)

  16. 30-th anniversary of nuclear power engineering in the USSR. Safety problems of nuclear engineering

    International Nuclear Information System (INIS)

    Kulov, E.V.; Sidorenko, V.A.; Kovalevich, O.M.

    1984-01-01

    The problems of safety and protection of personnel, population and environments from the NPP radiation effect in the Soviet Union in cases of possible accidents and in the course of normal operation are discussed. The problem of safety assurance includes scientific technological and organizational aspects. The necessity of accounting the ''national factor'' in the form of a possible NPP location, existing legislation experience gained, etc. is noted. The main trends in ensuring safety are a high quality of equipment manufacturing and assembly, quality control of equipment at all stages of operation, development and realization of efficient protective measures and devices, assurance of safety precautions at all stages of designing equipment production, NPP operation and maintenance, functioning of the system of state control of NPP safety assurance. The skill and training of personnel are necessary prerequisite for ensuring safety of nuclear power plants

  17. Environmental and safety problems of waste management in nuclear engineering

    International Nuclear Information System (INIS)

    Schwibach, J.; Jacobi, W.

    1976-01-01

    The environmental and safety problems which waste management in nuclear technology poses are discussed under the aspects of the disposal of radioactive waste by nuclear facilities and the safety of radioactive waste disposal. The release and global distribution of long-lived radionuclides such as tritium, Kr-85, C-14, I-129, and Pu-239 as well as the radiation exposure of the world population resulting thereof are investigated, the authors starting from a specific production rate of the nuclides released from nuclear facilities. Definitions of the terms 'dose commitment' and 'collective dose commitment' are given. Furthermore, local radiation exposure in reprocessing plants is investigated and compared with regional and global radiation exposure. A recommendation is made to take measures which would reduce the nuclides tritium, Kr-85, and C-14 in order to achieve considerably smaller collective doses. (HR/LN) [de

  18. Systems Engineering and Safety Issues in Scientific Facilities Subject to Ionizing Radiations

    Directory of Open Access Journals (Sweden)

    Pierre Bonnal

    2013-10-01

    Full Text Available The conception and development of large-scale scientific facilities emitting ionizing radiations rely more on project management practices in use in the process industry than on systems engineering practices. This paper aims to highlight possible reasons for this present situation and to propose some ways to enhance systems engineering so that the specific radiation safety requirements are considered and integrated in the approach. To do so, we have reviewed lessons learned from the management of large-scale scientific projects and more specifically that of the Large Hadron Collider project at CERN. It is shown that project management and systems engineering practices are complementary and can beneficially be assembled in an integrated and lean managerial framework that grants the appropriate amount of focus to safety and radiation safety aspects.

  19. Online probabilistic operational safety assessment of multi-mode engineering systems using Bayesian methods

    International Nuclear Information System (INIS)

    Lin, Yufei; Chen, Maoyin; Zhou, Donghua

    2013-01-01

    In the past decades, engineering systems become more and more complex, and generally work at different operational modes. Since incipient fault can lead to dangerous accidents, it is crucial to develop strategies for online operational safety assessment. However, the existing online assessment methods for multi-mode engineering systems commonly assume that samples are independent, which do not hold for practical cases. This paper proposes a probabilistic framework of online operational safety assessment of multi-mode engineering systems with sample dependency. To begin with, a Gaussian mixture model (GMM) is used to characterize multiple operating modes. Then, based on the definition of safety index (SI), the SI for one single mode is calculated. At last, the Bayesian method is presented to calculate the posterior probabilities belonging to each operating mode with sample dependency. The proposed assessment strategy is applied in two examples: one is the aircraft gas turbine, another is an industrial dryer. Both examples illustrate the efficiency of the proposed method

  20. Radiation safety system (RSS) backbones: Design, engineering, fabrication and installation

    International Nuclear Information System (INIS)

    Wilmarth, J.E.; Sturrock, J.C.; Gallegos, F.R.

    1998-01-01

    The Radiation Safety System (RSS) Backbones are part of an electrical/electronic/mechanical system insuring safe access and exclusion of personnel to areas at the Los Alamos Neutron Science Center (LANSCE) accelerator. The RSS Backbones control the safety fusible beam plugs which terminate transmission of accelerated ion beams in response to predefined conditions. Any beam or access fault of the backbone inputs will cause insertion of the beam plugs in the low energy beam transport. The Backbones serve the function of tying the beam plugs to the access control systems, beam spill monitoring systems and current-level limiting systems. In some ways the Backbones may be thought of as a spinal column with beam plugs at the head and nerve centers along the spinal column. The two Linac Backbone segments and experimental area segments form a continuous cable plant over 3,500 feet from beam plugs to the tip on the longest tail. The Backbones were installed in compliance with current safety standards, such as installation of the two segments in separate conduits or tray. Monitoring for ground-faults and input wiring verification was an added enhancement to the system. The system has the capability to be tested remotely

  1. Fundamentals of automotive and engine technology standard drives, hybrid drives, brakes, safety systems

    CERN Document Server

    2014-01-01

    Hybrid drives and the operation of hybrid vehicles are characteristic of contemporary automotive technology. Together with the electronic driver assistant systems, hybrid technology is of the greatest importance and both cannot be ignored by today’s car drivers. This technical reference book provides the reader with a firsthand comprehensive description of significant components of automotive technology. All texts are complemented by numerous detailed illustrations. Contents History of the automobile.- History of the Diesel engine.- Areas of use for Diesel engines.- Basic principles of the Diesel engine.- Basic principles of Diesel fuel-injection.- Basic principles of the gasoline engine.- Inductive ignition system.- Transmissions for motor vehicles.- Motor vehicle safety.- Basic principles of vehicle dynamics.- Car braking systems.- Vehicle electrical systems.- Overview of electrical and electronic systems in the vehicle.- Control of gasoline engines.- Control of Diesel engines.- Lighting technology.- Elec...

  2. Turboprop Engine Nacelle Optimization for Flight Increased Safety and Pollution Reduction

    Directory of Open Access Journals (Sweden)

    Cristian DOROBAT

    2018-03-01

    Full Text Available Commuter airplanes defined in CS-23 as being propeller driven, twin-engine, nineteen seats and maximum certified take-off weight of 8618 Kg had lately a special development due to advantages of turboprop engine compared with piston or jet engines. Nacelle optimization implies a sound and vibrations proof engine frame, engine fuel consumption reduction (through smaller nacelle drag and weight, better lift, better pressure recovery in air induction system, smaller drag of exhaust nozzles, engine cooling and nacelle ventilation more efficient, composite nacelle fairings with noise reduction properties, etc.. Nacelle aerodynamic experimental model, air induction experimental model and other nacelle experimental systems tested independently allow construction efficiency due to minimizing modifications on nacelle assembly and more safety in operation [1].

  3. Identification of features of electronic prescribing systems to support quality and safety in primary care using a modified Delphi process.

    Science.gov (United States)

    Sweidan, Michelle; Williamson, Margaret; Reeve, James F; Harvey, Ken; O'Neill, Jennifer A; Schattner, Peter; Snowdon, Teri

    2010-04-15

    Electronic prescribing is increasingly being used in primary care and in hospitals. Studies on the effects of e-prescribing systems have found evidence for both benefit and harm. The aim of this study was to identify features of e-prescribing software systems that support patient safety and quality of care and that are useful to the clinician and the patient, with a focus on improving the quality use of medicines. Software features were identified by a literature review, key informants and an expert group. A modified Delphi process was used with a 12-member multidisciplinary expert group to reach consensus on the expected impact of the features in four domains: patient safety, quality of care, usefulness to the clinician and usefulness to the patient. The setting was electronic prescribing in general practice in Australia. A list of 114 software features was developed. Most of the features relate to the recording and use of patient data, the medication selection process, prescribing decision support, monitoring drug therapy and clinical reports. The expert group rated 78 of the features (68%) as likely to have a high positive impact in at least one domain, 36 features (32%) as medium impact, and none as low or negative impact. Twenty seven features were rated as high positive impact across 3 or 4 domains including patient safety and quality of care. Ten features were considered "aspirational" because of a lack of agreed standards and/or suitable knowledge bases. This study defines features of e-prescribing software systems that are expected to support safety and quality, especially in relation to prescribing and use of medicines in general practice. The features could be used to develop software standards, and could be adapted if necessary for use in other settings and countries.

  4. Proposal of criteria for evaluation of engineering safety factors of VVER core parameters

    International Nuclear Information System (INIS)

    Shishkov, L.; Tsyganov, S.; Dementiev, V.

    2009-01-01

    The paper states that the regulatory documentation, as a rule, do not give explicit recommendations on formation techniques of engineering safety factors for design limited parameters of normal operation (K eng ). The AER countries use different approaches to K eng evaluation (sometimes even one country in relation of various power units). The paper suggests the development of uniform rules to be used in calculation of engineering safety factor for all VVER reactors. The paper presents principal problems that must be solved in the course of the discussion, and in the form of an exercise suggests the way of their solution. (authors)

  5. SAFETY ENGINEERING FOR THE RELATIVISTIC HEAVY ION COLLIDER AT THE BROOKHAVEN NATIONAL LABORATORY

    International Nuclear Information System (INIS)

    Musolino, S.V.; Kane, S.F.; Levesque, J.W.

    1999-01-01

    THERE ARE ONLY A FEW OTHER HIGH ENERGY PARTICLE ACCELERATORS LIKE RHIC IN THE WORLD. THEREFORE, THE DESIGNERS OF THE MACHINE DO NOT ALWAYS HAVE CONSENSUS DESIGN STANDARDS AND REGULATORY GUIDANCE AVAILABLE TO ESTABLISH THE ENGINEERING PARAMETERS FOR SAFETY. SOME OF THE AREAS WHERE STANDARDS ARE NOT AVAILABLE RELATE TO THE CRYOGENIC SYSTEM, CONTAINMENT OF LARGE VOLUMES OF FLAMMABLE GAS IN FRAGILE VESSELS IN THE EXPERIMENTAL APPARATUS AND MITIGATION OF A DESIGN BASIS ACCIDENT WITH A STORED PARTICLE BEAM. UNIQUE BUT EQUIVALENT SAFETY ENGINEERING MUST BE DETERMINED. SPECIAL DESIGN CRITERIA FOR PROMPT RADIATION WERE DEVELOPED TO PROVIDE GUIDANCE FOR THE DESIGN OF RADIATION SHIELDING

  6. Proposal of criteria for evaluation of engineering safety factors of WWER core parameters

    International Nuclear Information System (INIS)

    Shishkov, L.; Tsyganov, S.; Dementiev, V.

    2009-01-01

    The paper states that the regulatory documentation, as a rule, do not give explicit recommendations on formation techniques of engineering safety factors for design limited parameters of normal operation. The AER countries use different approaches to evaluation (sometimes even one country in relation of various power units). The paper suggests the development of uniform rules to be used in calculation of engineering safety factor for all WWER reactors. The paper presents principal problems that must be solved in the course of the discussion, and in the form of an exercise suggests the way of their solution. (Authors)

  7. Engineering nucleases for gene targeting: safety and regulatory considerations.

    Science.gov (United States)

    Pauwels, Katia; Podevin, Nancy; Breyer, Didier; Carroll, Dana; Herman, Philippe

    2014-01-25

    Nuclease-based gene targeting (NBGT) represents a significant breakthrough in targeted genome editing since it is applicable from single-celled protozoa to human, including several species of economic importance. Along with the fast progress in NBGT and the increasing availability of customized nucleases, more data are available about off-target effects associated with the use of this approach. We discuss how NBGT may offer a new perspective for genetic modification, we address some aspects crucial for a safety improvement of the corresponding techniques and we also briefly relate the use of NBGT applications and products to the regulatory oversight. Copyright © 2013 Elsevier B.V. All rights reserved.

  8. Safety considerations in the design of the Fusion Engineering Device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    The US Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris

  9. The engineering project and reliability research of the safety interlock slow control system in BESIII

    International Nuclear Information System (INIS)

    Zhang Yinhong; Zhao Jingwei; Li Xiaonan; Xie Xiaoxi; Gao Cuishan; Bai Jingzhi; Chen Xihui; Min Jian; Nie Zhendong

    2008-01-01

    The new safety interlock slow control system of BESIII is designed to ensure that the BESIII interior equipments and the accelerator control center to work in coordination, and to guarantee the safety of the operating staff and all the important equipments at the same time. This paper introduces the hardware and software design of safety interlock system from the engineering requirements angle, including a detailed research on the software implementation technique of the state machine on PLC and the reliability of the system. (authors)

  10. Optimization of the nuclear power engineering safety on the basis of social and economic parameters

    International Nuclear Information System (INIS)

    Kozlov, V.F.; Kuz'min, I.I.; Lystsov, V.N.; Amosova, T.V.; Makhutov, N.A.; Men'shikov, V.F.

    1995-01-01

    Principle of optimization of nuclear power engineering safety is presented on the basis of estimating the risks to the man's health with an account of peculiarities of socio-economic system and other types of economic activities in the region. Average expected duration of forthcoming life and costs of its prolongation serve as a unit for measuring the man's safety. It is shown that if the expenditures on NPP technical safety exceed the scientifically substantiated costs for this region with application of the above principle, than the risk for population will exceed the minimum achievable level. 8 refs., 2 figs., 1 tab

  11. Systems engineering applied to integrated safety management for high consequence facilities

    International Nuclear Information System (INIS)

    Barter, R; Morais, B.

    1998-01-01

    Integrated Safety Management is a concept that is being actively promoted by the U.S. Department of Energy as a means of assuring safe operation of its facilities. The concept involves the integration of safety precepts into work planning rather than adjusting for safe operations after defining the work activity. The system engineering techniques used to design an integrated safety management system for a high consequence research facility are described. An example is given to show how the concepts evolved with the system design

  12. Integrating RAMS engineering and management with the safety life cycle of IEC 61508

    International Nuclear Information System (INIS)

    Lundteigen, Mary Ann; Rausand, Marvin; Utne, Ingrid Bouwer

    2009-01-01

    This article outlines a new approach to reliability, availability, maintainability, and safety (RAMS) engineering and management. The new approach covers all phases of the new product development process and is aimed at producers of complex products like safety instrumented systems (SIS). The article discusses main RAMS requirements to a SIS and presents these requirements in a holistic perspective. The approach is based on a new life cycle model for product development and integrates this model into the safety life cycle of IEC 61508. A high integrity pressure protection system (HIPPS) for an offshore oil and gas application is used to illustrate the approach.

  13. Handling and feeding of biomass to pressurized reactors: safety engineering

    Energy Technology Data Exchange (ETDEWEB)

    Wilen, Carl; Rautalin, Aimo (Valtion Teknillinen Tutkimuskeskus, Espoo (Finland). Lab. of Fuel and Process Technology)

    1993-01-01

    There are rather few literature references to or experience of the feed of biomass into a pressurized space. Alternatives given in the literature usually concern handling and feeding technology for coal. Some screw- or piston-operated plug feeders and coal and concrete pump equipment have, however, also been tested with biomasses. Explosion characteristics of fuels and their susceptibility to spontaneous ignition have been studied at both atmospheric and elevated pressures. The maximum explosion pressure and maximum rate of pressure rise, being critical factors in the process design and in the choice of safety equipment, have been determined under these conditions. In pressurized processes, the maintenance of sufficient inertization in fuel-feed systems is an especially critical factor. Peat, bark, and forest residues were used as biofuels, and lignite was used as reference fuel. The results obtained with a dynamic method for spontaneous ignition were compared with experience obtained from the operation of a commercial pressurized peat gasifier of 140 MW. (author)

  14. Engineering solutions of traffic safety problems of road transport

    Directory of Open Access Journals (Sweden)

    M. Bogdevičius

    2004-02-01

    Full Text Available The authors of this paper focus on the simulation of the motor vehicle movement (taking into consideration motor vehicle dynamics, motor vehicle hydraulic brake system influence on motor vehicle movement, interaction between its wheels with road pavements, road guardrail characteristics, interaction between motor vehicle and road guardrail on a certain road section and propose their specific solution of this problem. The presented results, illustrating the motor vehicle movement trajectories (motor vehicle braking and interaction between motor vehicle and road guardrail at various initial conditions and at various certain pavement surface of the road section under investigation and work of a motor vehicle hydraulic brake system. Taking into consideration the presented general mathematical model and computer aided test results it is possible to investigate various road transport traffic situations as well as to investigate various transport traffic safety problems.

  15. Engineering Solutions to Enhance Traffic Safety Performance on Two-Lane Highways

    Directory of Open Access Journals (Sweden)

    Lina Wu

    2015-01-01

    Full Text Available Improving two-lane highway traffic safety conditions is of practical importance to the traffic system, which has attracted significant research attention within the last decade. Many cost-effective and proactive solutions such as low-cost treatments and roadway safety monitoring programs have been developed to enhance traffic safety performance under prevailing conditions. This study presents research perspectives achieved from the Highway Safety Enhancement Project (HSEP that assessed safety performance on two-lane highways in Beijing, China. Potential causal factors are identified based on proposed evaluation criteria, and primary countermeasures are developed against inferior driving conditions such as sharp curves, heavy gradients, continuous downgrades, poor sight distance, and poor clear zones. Six cost-effective engineering solutions were specifically implemented to improve two-lane highway safety conditions, including (1 traffic sign replacement, (2 repainting pavement markings, (3 roadside barrier installation, (4 intersection channelization, (5 drainage optimization, and (6 sight distance improvement. The effectiveness of these solutions was examined and evaluated based on Empirical Bayes (EB models. The results indicate that the proposed engineering solutions effectively improved traffic safety performance by significantly reducing crash occurrence risks and crash severities.

  16. A review of factors influencing underground construction quality and the performance of engineered features

    International Nuclear Information System (INIS)

    Talbot, R.; Moy, D.; Breeds, C.D.; Kostelec, C.M.; Emsley, S.J.; Shuttle, D.; Cutler, J.; Hedman, T.; Carlsson, A.

    1992-01-01

    This report describes research to identify and rank the factors that should be considered for an assessment of the radioactive containment capability of repository engineered features. Containment is to be achieved through the use of multiple barriers to radionuclide movement, including the waste, waste container, repository structures and backfill, and repository seals. A review of the literature is used to identify radionuclide mobilisation and transport processes for the anticipated environmental conditions, and to compile a data base of repository design features, functional requirements, and design objectives. For selected design components, the report identifies alternative designs and factors contributing to design, focusing on the requirements for a design process, design specifications, construction requirements, and quality assurance requirements for the both design and construction. Existing practice in large, underground civil construction projects is used as a part of comparison, in conjunction with reported performance assessments, to determine those areas in which the repository developer will need to provide additional data, justification, and documentation prior to construction to ensure that performance requirements can be achieved. (Author)

  17. Ethical issues in engineering design processes ; regulative frameworks for safety and sustainability

    NARCIS (Netherlands)

    Gorp, A. van

    2007-01-01

    The ways designers deal with ethical issues that arise in their consideration of safety and sustainability in engineering design processes are described. In the case studies, upon which this article is based, a difference can be seen between normal and radical design. Designers refer to regulative

  18. State-of-the-art WEB -technologies and ecological safety of nuclear power engineering facilities

    International Nuclear Information System (INIS)

    Batij, V.G.; Batij, E.V.; Rud'ko, V.M.; Kotlyarov, V.T.

    2004-01-01

    Prospects of web-technologies using in the field of improvement radiation safety level of nuclear power engineering facilities is seen. It is shown that application of such technologies will enable entirely using the data of all information systems of radiation control

  19. The Attitude of Civil Engineering Students towards Health and Safety Risk Management: A Case Study

    Science.gov (United States)

    Petersen, A. K.; Reynolds, J. H.; Ng, L. W. T.

    2008-01-01

    The highest rate of accidents and injuries in British industries has been reported by the construction industry during the past decade. Since then stakeholders have recognised that a possible solution would be to inculcate a good attitude towards health and safety risk management in undergraduate civil engineering students and construction…

  20. Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2016-05-15

    The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

  1. Resilience Engineering in Critical Long Term Aerospace Software Systems: A New Approach to Spacecraft Software Safety

    Science.gov (United States)

    Dulo, D. A.

    Safety critical software systems permeate spacecraft, and in a long term venture like a starship would be pervasive in every system of the spacecraft. Yet software failure today continues to plague both the systems and the organizations that develop them resulting in the loss of life, time, money, and valuable system platforms. A starship cannot afford this type of software failure in long journeys away from home. A single software failure could have catastrophic results for the spaceship and the crew onboard. This paper will offer a new approach to developing safe reliable software systems through focusing not on the traditional safety/reliability engineering paradigms but rather by focusing on a new paradigm: Resilience and Failure Obviation Engineering. The foremost objective of this approach is the obviation of failure, coupled with the ability of a software system to prevent or adapt to complex changing conditions in real time as a safety valve should failure occur to ensure safe system continuity. Through this approach, safety is ensured through foresight to anticipate failure and to adapt to risk in real time before failure occurs. In a starship, this type of software engineering is vital. Through software developed in a resilient manner, a starship would have reduced or eliminated software failure, and would have the ability to rapidly adapt should a software system become unstable or unsafe. As a result, long term software safety, reliability, and resilience would be present for a successful long term starship mission.

  2. Proceedings of the SRESA national conference on reliability and safety engineering

    International Nuclear Information System (INIS)

    Varde, P.V.; Vaishnavi, P.; Sujatha, S.; Valarmathi, A.

    2014-01-01

    The objective of this conference was to provide a forum for technical discussions on recent developments in the area of risk based approach and Prognostic Health Management of critical systems in decision making. The reliability and safety engineering methods are concerned with the way which the product fails, and the effects of failure is to understand how a product works and assures acceptable levels of safety. The reliability engineering addresses all the anticipated and possibly unanticipated causes of failure to ensure the occurrence of failure is prevented or minimized. The topics discussed in the conference were: Reliability in Engineering Design, Safety Assessment and Management, Reliability analysis and Assessment , Stochastic Petri nets for reliability Modeling, Dynamic Reliability, Reliability Prediction, Hardware Reliability, Software Reliability in Safety Critical Issues, Probabilistic Safety Assessment, Risk Informed Approach, Dynamic Models for Reliability Analysis, Reliability based Design and Analysis, Prognostics and Health Management, Remaining Useful Life (RUL), Human Reliability Modeling, Risk Based Applications, Hazard and Operability Study (HAZOP), Reliability in Network Security and Quality Assurance and Management etc. The papers relevant to INIS are indexed separately

  3. Safety Assessment of a Metal Cask under Aircraft Engine Crash

    Directory of Open Access Journals (Sweden)

    Sanghoon Lee

    2016-04-01

    Full Text Available The structural integrity of a dual-purpose metal cask currently under development by the Korea Radioactive Waste Agency (KORAD was evaluated, through numerical simulations and a model test, under high-speed missile impact reflecting targeted aircraft crash conditions. The impact conditions were carefully chosen through a survey on accident cases and recommendations from literature. In the impact scenario, a missile flying horizontally hits the top side of the cask, which is freestanding on a concrete pad, with a velocity of 150 m/s. A simplified missile simulating a commercial aircraft engine was designed from an impact load–time function available in literature. In the analyses, the dynamic behavior of the metal cask and the integrity of the containment boundary were assessed. The simulation results were compared with the test results for a 1:3 scale model. Although the dynamic behavior of the cask in the model test did not match exactly with the prediction from the numerical simulation, other structural responses, such as the acceleration and strain history during the impact, showed very good agreement. Moreover, the containment function of the cask survived the missile impact as expected from the numerical simulation. Thus, the procedure and methodology adopted in the structural numerical analyses were successfully validated.

  4. Education and training in nuclear engineering and safety

    International Nuclear Information System (INIS)

    Moons, F.; Safieh, J.; Giot, M.; Mavko, B.; Sehgal, Raj B.; Schaefer, A.; Van Goethem, G.; D'haeseleer, W.

    2007-01-01

    The need to preserve, enhance or strengthen nuclear knowledge is worldwide recognized since a couple of years. Within the 5th framework program the European Commission supports the European nuclear higher education network. The ENEN contract started on Jan 1, 2002 and lasts for 24 months. The Commission support for this 'accompanying measure' amounts to EUR 197 716. Based upon a year-long extensive exchange of views between the partners of ENEN, consisting of a representative cross section of nuclear academic institutions and research laboratories of the EU-25, a coherent and practicable concept for a European Master of Science in Nuclear Engineering has emerged. The concept is compatible with the Bologna philosophy of higher education for academic education in Europe. Pursuing the sustainability of the concept, the ENEN partners organized themselves in a non-profit-making association. Within the 6th framework program, the Commission services favourably evaluated the proposal: 'Nuclear European Platform of Training and University Organisations'. The objectives of the NEPTUNO co-ordination action are to establish a fair dialogue and a strong interaction between the academic and the industrial world and to bring all nuclear education and training activities under a common strategy of the ENEN type. The present proposal schedules for 18 months and the Commission earmarked a financial contribution of EUR 830 619. (author)

  5. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    International Nuclear Information System (INIS)

    JANICEK, G.P.

    2000-01-01

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance

  6. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    Energy Technology Data Exchange (ETDEWEB)

    JANICEK, G.P.

    2000-06-08

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance.

  7. The Increase of Operational Safety of Ships by Improving Diagnostic Methods for Marine Diesel Engine

    Directory of Open Access Journals (Sweden)

    Kazimierz Witkowski

    2017-06-01

    Full Text Available This article shows the importance of the diagnostic improvement methods of marine engines to boost the economy and safety of operation of marine cargo ships. The need to implement effective diagnostic methods is justified by presenting statistical data of marine diesel engines failure and the cost of their operation. Based on the own research has been proven, for the chosen example, that indicator diagrams and analysis of indicated parameters have limited utility in the diagnosis of damages of marine engine, although this is a method commonly used in operational practice. To achieve greater diagnostic effectiveness, when, based on indicator diagrams, are calculated and then the characteristics of heat release is analyzed - net of heat release characteristics and the intensity of the heat release, it was demonstrated. This procedure is particularly effective in the diagnosis of damage of injection system components marine diesel engine.

  8. Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

    International Nuclear Information System (INIS)

    Bess, J.D.; Briggs, J.B.; Garcia, A.S.

    2011-01-01

    One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

  9. Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; J. B. Briggs; A. S. Garcia

    2011-09-01

    One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

  10. Study of Machine-Learning Classifier and Feature Set Selection for Intent Classification of Korean Tweets about Food Safety

    Directory of Open Access Journals (Sweden)

    Yeom, Ha-Neul

    2014-09-01

    Full Text Available In recent years, several studies have proposed making use of the Twitter micro-blogging service to track various trends in online media and discussion. In this study, we specifically examine the use of Twitter to track discussions of food safety in the Korean language. Given the irregularity of keyword use in most tweets, we focus on optimistic machine-learning and feature set selection to classify collected tweets. We build the classifier model using Naive Bayes & Naive Bayes Multinomial, Support Vector Machine, and Decision Tree Algorithms, all of which show good performance. To select an optimum feature set, we construct a basic feature set as a standard for performance comparison, so that further test feature sets can be evaluated. Experiments show that precision and F-measure performance are best when using a Naive Bayes Multinomial classifier model with a test feature set defined by extracting Substantive, Predicate, Modifier, and Interjection parts of speech.

  11. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

    JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

  12. Safety culture for engineering companies. Licensing and design bases for Cofrentes NPP

    International Nuclear Information System (INIS)

    Nhorte Gomez, M.D.

    1994-01-01

    Safety culture must be given higher priority by all organisations. It must not be considered a separate concept, attributable to just one particular organisation, or a single responsible party. It is important to apply this criterion throughout the different phases of a nuclear power plant project (design, construction, commissioning and operation) without becoming isolated or dissociated. Nevertheless, it is absolutely essential to apply and consider it during operation, so to ensure highest possible safety standards. Consideration must also be given to the interfaces and interconnections between the different parties involved in the project (Owner of the NPP, Main Engineering Company, Main Supplier, Regulatory Body, etc) to build a SAFETY CULTURE in a collective and effective way. In applying the safety culture, an engineering company emphasises the following concepts: - Personal dedication and sense of responsibility in all those involved in any activity related to the safety of Nuclear Power Plants. - Clearly defined and readily accessible areas of responsibility and channels of communication - Strict adherence to procedures - Internal review of activities (Design review) (Author)

  13. Idaho National Engineering Laboratory (INEL) Environmental Restoration (ER) Program Baseline Safety Analysis File (BSAF)

    International Nuclear Information System (INIS)

    1995-09-01

    The Baseline Safety Analysis File (BSAF) is a facility safety reference document for the Idaho National Engineering Laboratory (INEL) environmental restoration activities. The BSAF contains information and guidance for safety analysis documentation required by the U.S. Department of Energy (DOE) for environmental restoration (ER) activities, including: Characterization of potentially contaminated sites. Remedial investigations to identify and remedial actions to clean up existing and potential releases from inactive waste sites Decontamination and dismantlement of surplus facilities. The information is INEL-specific and is in the format required by DOE-EM-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports. An author of safety analysis documentation need only write information concerning that activity and refer to BSAF for further information or copy applicable chapters and sections. The information and guidance provided are suitable for: sm-bullet Nuclear facilities (DOE Order 5480-23, Nuclear Safety Analysis Reports) with hazards that meet the Category 3 threshold (DOE-STD-1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports) sm-bullet Radiological facilities (DOE-EM-STD-5502-94, Hazard Baseline Documentation) Nonnuclear facilities (DOE-EM-STD-5502-94) that are classified as open-quotes lowclose quotes hazard facilities (DOE Order 5481.1B, Safety Analysis and Review System). Additionally, the BSAF could be used as an information source for Health and Safety Plans and for Safety Analysis Reports (SARs) for nuclear facilities with hazards equal to or greater than the Category 2 thresholds, or for nonnuclear facilities with open-quotes moderateclose quotes or open-quotes highclose quotes hazard classifications

  14. Idaho National Engineering Laboratory (INEL) Environmental Restoration (ER) Program Baseline Safety Analysis File (BSAF)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The Baseline Safety Analysis File (BSAF) is a facility safety reference document for the Idaho National Engineering Laboratory (INEL) environmental restoration activities. The BSAF contains information and guidance for safety analysis documentation required by the U.S. Department of Energy (DOE) for environmental restoration (ER) activities, including: Characterization of potentially contaminated sites. Remedial investigations to identify and remedial actions to clean up existing and potential releases from inactive waste sites Decontamination and dismantlement of surplus facilities. The information is INEL-specific and is in the format required by DOE-EM-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports. An author of safety analysis documentation need only write information concerning that activity and refer to BSAF for further information or copy applicable chapters and sections. The information and guidance provided are suitable for: {sm_bullet} Nuclear facilities (DOE Order 5480-23, Nuclear Safety Analysis Reports) with hazards that meet the Category 3 threshold (DOE-STD-1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports) {sm_bullet} Radiological facilities (DOE-EM-STD-5502-94, Hazard Baseline Documentation) Nonnuclear facilities (DOE-EM-STD-5502-94) that are classified as {open_quotes}low{close_quotes} hazard facilities (DOE Order 5481.1B, Safety Analysis and Review System). Additionally, the BSAF could be used as an information source for Health and Safety Plans and for Safety Analysis Reports (SARs) for nuclear facilities with hazards equal to or greater than the Category 2 thresholds, or for nonnuclear facilities with {open_quotes}moderate{close_quotes} or {open_quotes}high{close_quotes} hazard classifications.

  15. System Coordination of Survivability and Safety of Complex Engineering Objects Operation

    Directory of Open Access Journals (Sweden)

    Nataliya Pankratova

    2014-11-01

    Full Text Available A system strategy to estimation the guaranteed survivability and safety of complex engineering objects (CEO operation is proposed. The principles that underlie the strategy of the guaranteed safety of CEO operation provide a flexible approach to timely detection, recognition, forecast, and system diagnostics of risk factors and situations, to formulation and implementation of a rational decision in a practicable time within an unremovable time constraint. Implementation of the proposed strategy is shown on example of diagnostics of electromobile-refrigerator functioning in real mode.

  16. Engineered barrier systems (EBS) in the context of the entire safety case

    International Nuclear Information System (INIS)

    2003-01-01

    A joint NEA-EC workshop entitled 'Engineered Barrier Systems (EBS) in the Context of the Entire Safety Case' was organised in Oxford on 25-27 September 2002 and hosted by United Kingdom Nirex Limited. The main objectives of the workshop were to provide a status report on engineered barrier systems in various national radioactive waste management programmes considering deep geological disposal; to establish the value to member countries of a project on EBS; and to define such a project's scope, timetable and modus operandi. This report presents the outcomes of this workshop. (author)

  17. Engineered Barrier Systems (EBS) in the Context of the Entire Safety Case

    International Nuclear Information System (INIS)

    2005-01-01

    A joint NEA-EC workshop entitled ''Engineered Barrier Systems (EBS) in the Context of the Entire Safety Case'' was organised in Oxford on 25-27 September 2002 and hosted by United Kingdom Nirex Limited. The main objectives of the workshop were to provide a status report on engineered barrier systems in various national radioactive waste management programmes considering deep geological disposal; to establish the value to member countries of a project on EBS; and to define such a project scope, timetable and modus operandi. This report presents the outcomes of this workshop. (author)

  18. Assessment of shaft safety and management system of controlling engineering information

    Energy Technology Data Exchange (ETDEWEB)

    Liu Rui-xin; Xu Yan-chun [Yanzhou Mining Group Ltd., Zoucheng (China)

    2008-02-15

    Evaluating shaft safety and establishing a system for controlling engineering information is very important because more than 90 shafts in thick alluvial areas suddenly have shaft wall fracturing or breaking problems and there are more than a few hundred shafts of similar geologic conditions. Taking shaft control in the Yangzhou Coal Mining Group as an example, an assessment and management system and related software were established. This system includes basic information of the mine, measurement results and analysis, and functions of empirical and theoretical forecasting and finite element analysis, which are confirmed to be very effective for guiding shaft well control engineering in practice. 8 refs., 3 figs., 2 tabs.

  19. A systems engineering approach to implementation of safety management systems in the Norwegian fishing fleet

    International Nuclear Information System (INIS)

    McGuinness, Edgar; Utne, Ingrid B.

    2014-01-01

    The fishing industry is plagued by a long history of fatality and injury occurrence. Commercial fishing is hence recognized as the most dangerous and difficult of professional callings, in all jurisdictions. Fishing vessels have their own unique set of hazards, a myriad collection of complex occupational accident potentials, barely controlled, co-existing in a perilous work environment. The work in this article is directed by the Norwegian Systematic Health, Environmental and Safety Activities in Enterprises (1997) (Internal Control Regulations [1]), the ISM Code [2] for vessels and their recent applicability to the fishing fleet of Norway. Both safety management works place requirements on the vessel operators and crew to actively manage safety as an on-going concern. The application of these safety management system (SMS) control documents to fishing vessels is just the latest instalment in a continual drive to improve safety in this sector. The difficulty is that there has been no previous systematic approach to safety within the fishing fleet. This article uses the tenants of systems engineering to determine the requirements for such a SMS, detailing the limiting factors and restrictive issues of this complex operating environment. - Highlights: • Systems engineer is applied as a tool for determining requirements for design and construction of a safety management system (SMS). • Outlining a simplistic format, identifying, designingand facilitating improvement opportunities in the conduction and application of SMS’s on fishing vessels. • Knowledge provision is a key requirement of management systems, through provision of understanding, detail orientation and applicable skills for realization. • Outlining, what is to be done and how it is to be completed to accomplish compliance with pertinent legislative requirements. • Promoting a combination of documentation and communication arrangements by which the actionsnecessary for management can be

  20. Safety assessment of Novi Han radioactive waste repository - features, problems, results and perspectives

    International Nuclear Information System (INIS)

    Mateeva, M.

    2000-01-01

    This paper summarizes the work done and the achievements reached in the Novi Han radioactive waste repository safety assessment within the IAEA Model Project 'Increasing the safety of Novi Han radioactive waste repository BUL 4/005'. The overall safety assessment has a wide context, but the work reported here relates only to some details and results concerning the development and implementation of the appropriate methodology approach, model and computer code used for the calculations. Different steps and procedures are included for a better practical understanding of the obtained results during the safety assessment performance. The methodology approach is widely based on an international experience in safety analysis and implemented for evaluation computer code AMBER, which is one of the recommended from the safety assessments experts. (author)

  1. Technology development of maintenance optimization and reliability analysis for safety features in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Choi, Seong Soo; Lee, Dong Gue; Kim, Young Il

    1999-12-01

    The reliability data management system (RDMS) for safety systems of PHWR type plants has been developed and utilized in the reliability analysis of the special safety systems of Wolsong Unit 1,2 with plant overhaul period lengthened. The RDMS is developed for the periodic efficient reliability analysis of the safety systems of Wolsong Unit 1,2. In addition, this system provides the function of analyzing the effects on safety system unavailability if the test period of a test procedure changes as well as the function of optimizing the test periods of safety-related test procedures. The RDMS can be utilized in handling the requests of the regulatory institute actively with regard to the reliability validation of safety systems. (author)

  2. Anomaly Analysis: NASA's Engineering and Safety Center Checks Recurring Shuttle Glitches

    Science.gov (United States)

    Morring, Frank, Jr.

    2004-01-01

    The NASA Engineering and Safety Center (NESC), set up in the wake of the Columbia accident to backstop engineers in the space shuttle program, is reviewing hundreds of recurring anomalies that the program had determined don't affect flight safety to see if in fact they might. The NESC is expanding its support to other programs across the agency, as well. The effort, which will later extend to the International Space Station (ISS), is a principal part of the attempt to overcome the normalization of deviance--a situation in which organizations proceeded as if nothing was wrong in the face of evidence that something was wrong--cited by sociologist Diane Vaughn as contributing to both space shuttle disasters.

  3. Human factors engineering design review acceptance criteria for the safety parameter display

    International Nuclear Information System (INIS)

    McGevna, V.; Peterson, L.R.

    1981-01-01

    This report contains human factors engineering design review acceptance criteria developed by the Human Factors Engineering Branch (HFEB) of the Nuclear Regulatory Commission (NRC) to use in evaluating designs of the Safety Parameter Display System (SPDS). These criteria were developed in response to the functional design criteria for the SPDS defined in NUREG-0696, Functional Criteria for Emergency Response Facilities. The purpose of this report is to identify design review acceptance criteria for the SPDS installed in the control room of a nuclear power plant. Use of computer driven cathode ray tube (CRT) displays is anticipated. General acceptance criteria for displays of plant safety status information by the SPDS are developed. In addition, specific SPDS review criteria corresponding to the SPDS functional criteria specified in NUREG-0696 are established

  4. Definition and Means of Maintaining the Criticality Prevention Design Features Portion of the PFP Safety Envelope

    International Nuclear Information System (INIS)

    RAMBLE, A.L.

    2000-01-01

    The purpose of this document is to record the technical evaluation of the Operational Safety Requirements described in the Plutonium Finishing Plant Final (PFP) Operational Safety Requirements, WHC-SD-CP-OSR-010. Rev. 0-N , Section 3.1.1, ''Criticality Prevention System.'' This document, with its appendices, provides the following: (1) The results of a review of Criticality Safety Analysis Reports (CSAR), later called Criticality Safety Evaluation Reports (CSER), and Criticality Prevention Specifications (CPS) to determine which equipment or components analyzed in the CSER or CPS are considered as one of the two unlikely, independent, and concurrent changes before a criticality accident is possible. (2) Evaluations of equipment or components to determine the safety boundary for the system (Section 4). (3) A list of essential drawings that show the safety system or component (Appendix A). (4) A list of the safety envelope (SE) equipment (Appendix B). (5) Functional requirements for the individual safety envelope equipment (Sections 3 and 4). (6) A list of the operational and surveillance procedures necessary to maintain the system equipment within the safety envelope (Section 5)

  5. NASA Engineering and Safety Center (NESC) Enhanced Melamine (ML) Foam Acoustic Test (NEMFAT)

    Science.gov (United States)

    McNelis, Anne M.; Hughes, William O.; McNelis, Mark E.

    2014-01-01

    The NASA Engineering and Safety Center (NESC) funded a proposal to achieve initial basic acoustic characterization of ML (melamine) foam, which could serve as a starting point for a future, more comprehensive acoustic test program for ML foam. A project plan was developed and implemented to obtain acoustic test data for both normal and enhanced ML foam. This project became known as the NESC Enhanced Melamine Foam Acoustic Test (NEMFAT). This document contains the outcome of the NEMFAT project.

  6. Overheads, Safety Analysis and Engineering FY 1995 Site Support Program Plan WBS 6.3.5

    Energy Technology Data Exchange (ETDEWEB)

    DiVincenzo, E.P.

    1994-09-27

    The Safety Analysis & Engineering (SA&E) department provides core competency for safety analysis and risk documentation that supports achievement of the goals and mission as described in the Hanford Mission Plan, Volume I, Site Guidance (DOE-RL 1993). SA&E operations are integrated into the programs that plan and conduct safe waste management, environmental restoration, and operational activities. SA&E personnel are key members of task teams assigned to eliminate urgent risks and inherent threats that exist at the Hanford Site. Key to ensuring protection of public health and safety, and that of onsite workers, are the products and services provided by the department. SA&E will continue to provide a leadership role throughout the DOE complex with innovative, cost-effective approaches to ensuring safety during environmental cleanup operations. The SA&E mission is to provide support to direct program operations through safety analysis and risk documentation and to maintain an infrastructure responsive to the evolutionary climate at the Hanford Site. SA&E will maintain the appropriate skills mix necessary to fulfill the customers need to conduct all operations in a safe and cost-effective manner while ensuring the safety of the public and the onsite worker.

  7. Safety risk assessment for vertical concrete formwork activities in civil engineering construction.

    Science.gov (United States)

    López-Arquillos, Antonio; Rubio-Romero, Juan Carlos; Gibb, Alistair G F; Gambatese, John A

    2014-01-01

    The construction sector has one of the worst occupational health and safety records in Europe. Of all construction tasks, formwork activities are associated with a high frequency of accidents and injuries. This paper presents an investigation of the activities and related safety risks present in vertical formwork for in-situ concrete construction in the civil engineering sector. Using the methodology of staticized groups, twelve activities and ten safety risks were identified and validated by experts. Every safety risk identified in this manner was quantified for each activity using binary methodology according to the frequency and severity scales developed in prior research. A panel of experts was selected according to the relevant literature on staticized groups. The results obtained show that the activities with the highest risk in vertical formwork tasks are: Plumbing and leveling of forms, cutting of material, handling materials with cranes, and climbing or descending ladders. The most dangerous health and safety risks detected were falls from height, cutting and overexertion. The research findings provide construction practitioners with further evidence of the hazardous activities associated with concrete formwork construction and a starting point for targeting worker health and safety programmes.

  8. Overheads, Safety Analysis and Engineering FY 1995 Site Support Program Plan WBS 6.3.5

    International Nuclear Information System (INIS)

    DiVincenzo, E.P.

    1994-01-01

    The Safety Analysis ampersand Engineering (SA ampersand E) department provides core competency for safety analysis and risk documentation that supports achievement of the goals and mission as described in the Hanford Mission Plan, Volume I, Site Guidance (DOE-RL 1993). SA ampersand E operations are integrated into the programs that plan and conduct safe waste management, environmental restoration, and operational activities. SA ampersand E personnel are key members of task teams assigned to eliminate urgent risks and inherent threats that exist at the Hanford Site. Key to ensuring protection of public health and safety, and that of onsite workers, are the products and services provided by the department. SA ampersand E will continue to provide a leadership role throughout the DOE complex with innovative, cost-effective approaches to ensuring safety during environmental cleanup operations. The SA ampersand E mission is to provide support to direct program operations through safety analysis and risk documentation and to maintain an infrastructure responsive to the evolutionary climate at the Hanford Site. SA ampersand E will maintain the appropriate skills mix necessary to fulfill the customers need to conduct all operations in a safe and cost-effective manner while ensuring the safety of the public and the onsite worker

  9. Investigation of ability to guess safety signs based on cognitive features in one of the petrochemical industries

    Directory of Open Access Journals (Sweden)

    G. A. Shirali

    2015-07-01

    .Conclusion: According to results of this study, use of principles of ergonomic design of signs and training are necessary to promote the ability to guess the safety signs to the minimum available standards. Therefore, it is possible to balance cognitive features especially “familiarity”, with the lowest score, and “meaningfulness” and “semantic closeness”, with the highest influential relationship with the ability to guess of signs. The developed regression model for this industry can be used to predict the ability to guess of safety signs in future studies

  10. 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

  11. Electrochemical performance and safety features of high-safety lithium ion battery using novel branched additive for internal short protection

    International Nuclear Information System (INIS)

    Li Yuhan; Lee, Meng-Lun; Wang Fuming; Yang, Chang-Rung; Chu, Peter P.J.; Yau, Shueh-Lin; Pan, Jing-Pin

    2012-01-01

    Highlights: ► N-phenylmaleimide-containing branched oligomer has been employed as an additive in lithium cells. ► The branched oligomer additive enhances safety and cycling performance of Li ion battery. ► The highest temperature of branched oligomer-containing battery was only 85 °C in the nail penetration test. - Abstract: In this study, we have investigated N-phenylmaleimide/bismaleimide-containing branched oligomer (BO1) as additive in Li-ion batteries to increase the safety performance by reducing the probability of batteries suffering an internal short circuit. In the nail penetration test, a LiCoO 2 /MCMB full battery with N-phenylmaleimide/bismaleimide-containing branched oligomer (BO1) showed a significant improvement in thermal stability and was able to restrain the temperature of the battery at about 85 °C. Furthermore, we found that N-phenylmaleimide/bismaleimide-containing branched oligomer (BO1) contained battery revealed better cycling and electrochemical performance, compared with the battery with bismaleimide-containing branched oligomer (BO3) in the electrolyte. The improvement might result from the favorable ionic conductivity, Li ion mobility and lower resistance in the battery. This additive can meet the cycling performance and safety requirements for Li-ion batteries.

  12. Constructing engineers through practice: Gendered features of learning and identity development

    Science.gov (United States)

    Tonso, Karen L.

    How do women and men student engineers develop an engineering identity (a sense of belonging, or not), while practicing "actual" engineering? What are the influences of gender, learning and knowledge, relations of power, and conceptions of equality on cultural identity development? I studied these issues in reform-minded engineering design classes, courses organized around teaching students communications, teamwork, and practical engineering. Engineering-student cultural identity categories revealed a status hierarchy, predicated on meeting "academic" criteria for excellence, and the almost total exclusion of women. While working as an engineering colleague on five student teams (three first-year and two senior) and attending their design classes, I documented how cultural identities were made evident and constructed in students' practical engineering. Design projects promoted linking academic knowledge with real-world situations, sharing responsibilities and trusting colleagues, communicating engineering knowledge to technical and non-technical members of business communities, and addressing gaps in students' knowledge. With a curriculum analysis and survey of students' perceptions of the differences between design and conventional courses, I embedded the design classes in the wider campus and found that: (1) Engineering education conferred prestige, power, and well-paying jobs on students who performed "academic" engineering, while failing to adequately encourage "actual" engineering practices. High-status student engineers were the least likely to perform "actual" engineering in design teams. (2) Engineering education advanced an ideology that encouraged its practitioners to consider men's privilege and women's invisibility normal. By making "acting like men act" the standards to which engineering students must conform, women learned to put up with oppressive treatment. Women's accepting their own mistreatment and hiding their womanhood became a condition of

  13. Safety research experiment facilities, Idaho National Engineering Laboratory, Idaho. Final environmental impact statement

    International Nuclear Information System (INIS)

    Liverman, J.L.

    1977-09-01

    This environmental statement was prepared for the Safety Research Experiment Facilities (SAREF) Project. The purpose of the proposed project is to modify some existing facilities and provide a new test facility at the Idaho National Engineering Laboratory (INEL) for conducting fast breeder reactor (FBR) safety experiments. The SAREF Project proposal has been developed after an extensive study which identified the FBR safety research needs requiring in-reactor experiments and which evaluated the capability of various existing and new facilities to meet these needs. The proposed facilities provide for the in-reactor testing of large bundles of prototypical FBR fuel elements under a wide variety of conditions, ranging from those abnormal operating conditions which might be expected to occur during the life of an FBR power plant to the extremely low probability, hypothetical accidents used in the evaluation of some design options and in the assessment of the long-term potential risk associated with wide-acale deployment of the FBR

  14. Development of Risk Assessment Matrix for NASA Engineering and Safety Center

    Science.gov (United States)

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

    2004-01-01

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

  15. Nuclear technology and reactor safety engineering. The situation ten years after the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1996-01-01

    Ten years ago, on April 26, 1986 the most serious accident ever in the history of nuclear tgechnology worldwide happened in unit 4 of the nuclear power plant in Chernobyl in the Ukraine, this accident unveiling to the world at large that the Soviet reactor design lines are bearing unthought of safety engineering deficits. The dimensions of this reactor accident on site, and the radioactive fallout spreading far and wide to many countries in Europe, vividly nourished the concern of great parts of the population in the Western world about the safety of nuclear technology, and re-instigated debates about the risks involved and their justification. Now that ten years have elapsed since the accident, it is appropriate to strike a balance and analyse the situation today. The number of nuclear power plants operating worldwide has been growing in the last few years and this trend will continue, primarily due to developments in Asia. The Chernobyl reactor accident has pushed the international dimension of reactor safety to the foreground. Thus the Western world had reason enough to commit itself to enhancing the engineered safety of reactors in East Europe. The article analyses some of the major developments and activities to date and shows future perspectives. (orig.) [de

  16. Purpose, Principles, and Challenges of the NASA Engineering and Safety Center

    Science.gov (United States)

    Gilbert, Michael G.

    2016-01-01

    NASA formed the NASA Engineering and Safety Center in 2003 following the Space Shuttle Columbia accident. It is an Agency level, program-independent engineering resource supporting NASA's missions, programs, and projects. It functions to identify, resolve, and communicate engineering issues, risks, and, particularly, alternative technical opinions, to NASA senior management. The goal is to help ensure fully informed, risk-based programmatic and operational decision-making processes. To date, the NASA Engineering and Safety Center (NESC) has conducted or is actively working over 600 technical studies and projects, spread across all NASA Mission Directorates, and for various other U.S. Government and non-governmental agencies and organizations. Since inception, NESC human spaceflight related activities, in particular, have transitioned from Shuttle Return-to-Flight and completion of the International Space Station (ISS) to ISS operations and Orion Multi-purpose Crew Vehicle (MPCV), Space Launch System (SLS), and Commercial Crew Program (CCP) vehicle design, integration, test, and certification. This transition has changed the character of NESC studies. For these development programs, the NESC must operate in a broader, system-level design and certification context as compared to the reactive, time-critical, hardware specific nature of flight operations support.

  17. Most common road safety engineering deficiencies in South Eastern Europe as a part of safe system approach

    Science.gov (United States)

    Jovanov, D.; Vollpracht, H. J.; Beles, H.; Popa, V.; Tolea, B. A.

    2017-10-01

    Most common road safety engineering deficiencies identified by the authors in South Eastern Europe, including Romania, have been collected together and presented in this paper as a part of road safety unbreakably connected to the safe system approach (driver-vehicle-road). In different South Eastern Europe countries Road Safety Audit (RSA), Road Safety Inspection (RSI), as well as Black Spot Management (BSM) was introduced and practical implementation experience enabled the authors to analyze the road safety problems. Typical road safety engineering deficiencies have been presented in 8 different subsections, based on PIARC (World Road Association) RSA approach. This paper presents collected common road safety problems with relevant illustrations (real pictures) with associated accident risks.

  18. Completion of the VVER 440/213 NPP Mochovce incorporation enhanced safety features

    International Nuclear Information System (INIS)

    Charbonneau, S.; Eckert, G.

    1996-01-01

    The cooperation between the western countries and the countries of ex-eastern block in the field of nuclear safety is recent and still limited. The main reasons for this situation are limited or non existent capabilities of these countries for financing as well as non acceptable legal conditions concerning the third party nuclear liability in this part of Europe. Nevertheless, Framatome and Siemens associated in the consortium named EUCOM, have signed in April 1996 the contract of about 100 million US dollars with Slovak electricity company (SLOVENSKE ELEKTRARNE-SE) for upgrading the Units 1 and 2 of Mochovce Nuclear Power Plant according to the western safety standards. This is the first important project involving west-european companies in the modernisation of Russian type of pressurized water reactor (VVER 440/213). The consortium will cooperate with other partners involved in the project: Slovak, Czech and Russian. The financing of the project will be provided mainly form Slovak and Czech sources. The safety upgrading will be financed through French and German buyer credits. French company Electricite de France (EDF) will be the consultant for SE. The safety upgrading measures have been elaborated taking into account the recommendation of Vienna International Atomic Energy Agency (IAEA) and the evaluation of the safety realised by RISKAUDIT, the common organization of German and French safety authorities (GSR and IPSN). Hence all guaranties have been taken to fulfil the western safety criteria for Nuclear Power Plant Mochovce. (author)

  19. Development of feature extraction analysis for a multi-functional optical profiling device applied to field engineering applications

    Science.gov (United States)

    Han, Xu; Xie, Guangping; Laflen, Brandon; Jia, Ming; Song, Guiju; Harding, Kevin G.

    2015-05-01

    In the real application environment of field engineering, a large variety of metrology tools are required by the technician to inspect part profile features. However, some of these tools are burdensome and only address a sole application or measurement. In other cases, standard tools lack the capability of accessing irregular profile features. Customers of field engineering want the next generation metrology devices to have the ability to replace the many current tools with one single device. This paper will describe a method based on the ring optical gage concept to the measurement of numerous kinds of profile features useful for the field technician. The ring optical system is composed of a collimated laser, a conical mirror and a CCD camera. To be useful for a wide range of applications, the ring optical system requires profile feature extraction algorithms and data manipulation directed toward real world applications in field operation. The paper will discuss such practical applications as measuring the non-ideal round hole with both off-centered and oblique axes. The algorithms needed to analyze other features such as measuring the width of gaps, radius of transition fillets, fall of step surfaces, and surface parallelism will also be discussed in this paper. With the assistance of image processing and geometric algorithms, these features can be extracted with a reasonable performance. Tailoring the feature extraction analysis to this specific gage offers the potential for a wider application base beyond simple inner diameter measurements. The paper will present experimental results that are compared with standard gages to prove the performance and feasibility of the analysis in real world field engineering. Potential accuracy improvement methods, a new dual ring design and future work will be discussed at the end of this paper.

  20. 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

  1. 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

  2. Knowledge, attitude and practices for design for safety: A study on civil & structural engineers.

    Science.gov (United States)

    Goh, Yang Miang; Chua, Sijie

    2016-08-01

    Design for safety (DfS) (also known as prevention through design, safe design and Construction (Design and Management)) promotes early consideration of safety and health hazards during the design phase of a construction project. With early intervention, hazards can be more effectively eliminated or controlled leading to safer worksites and construction processes. DfS is practiced in many countries, including Australia, the UK, and Singapore. In Singapore, the Manpower Ministry enacted the DfS Regulations in July 2015, which will be enforced from August 2016 onwards. Due to the critical role of civil and structural (C&S) engineers during design and construction, the DfS knowledge, attitude and practices (KAP) of C&S engineers have significant impact on the successful implementation of DfS. Thus, this study aims to explore the DfS KAP of C&S engineers so as to guide further research in measuring and improving DfS KAP of designers. During the study, it was found that there is a lack of KAP studies in construction management. Therefore, this study also aims to provide useful lessons for future applications of the KAP framework in construction management research. A questionnaire was developed to assess the DfS KAP of C&S engineers. The responses provided by 43 C&S engineers were analyzed. In addition, interviews with experienced construction professionals were carried out to further understand perceptions of DfS and related issues. The results suggest that C&S engineers are supportive of DfS, but the level of DfS knowledge and practices need to be improved. More DfS guidelines and training should be made available to the engineers. To ensure that DfS can be implemented successfully, there is a need to study the contractual arrangements between clients and designers and the effectiveness of different implementation approaches for the DfS process. The questionnaire and findings in this study provided the foundation for a baseline survey with larger sample size, which is

  3. The impact of the European health and safety directives on engineering in higher education

    Science.gov (United States)

    Crisp, Alan Roy

    This thesis examines the effect that six sets of Health and Safety legislation introduced in 1993 have had on working practices at the University, particularly within the Engineering Departments. The legislation, collectively known colloquially as "the six pack", had much in common with extant United Kingdom (UK) law but, because it emanated from the European Union (EU), it appears to be viewed in the UK as unduly restrictive and time consuming. Much of the thesis is therefore devoted to examining this suspicion in which the EU and its legislation is held by UK employers and employees. The thesis begins by examining the general background and recent history of the EU, before going on to look in greater detail at the development of Health and Safety legislation in particular. The area of interest is then further narrowed to look at the impact of this legislation on Higher Education Institutions by comparing recent accident statistics with those for industry and commerce. The main outcome of this section is that Higher Education has a similar accident profile by 'type' to industry and commerce and therefore would act in a similar manner when implementing the legislation. It is argued that industry and commerce can benefit from this similarity by emulating two case studies at the University where legislation is applied to some engineering equipment and procedures. These are described in detail and the point is made that safety is an approach that pervades all stages of an engineering process, commencing with the design or ordering of equipment. This is reinforced with the results of a primary survey of purchasing at similar institutions with regards to observance of current safety practices. It is concluded that suspicion of the "six pack" legislation is largely the result of overloading of those people responsible for safety by the arrival of a plethora of legislation all at once. Ironically this overloading appears to have influenced safety officers to pay attention

  4. In vitro dosimetry modeling will be a critical step toward efficient assessment of engineered nanomaterials for environmental health and safety

    Science.gov (United States)

    Presentation Description: The development and application of engineered nanomaterials (ENM) into commercial and consumer products is far outpacing the ability of traditional approaches to evaluate the potential implications for environmental health and safety. This problem recen...

  5. Optimized maintenance concept of safety relevant valves related to ageing management features in nuclear power plants

    International Nuclear Information System (INIS)

    Koring, R.

    2007-01-01

    This paper presents the existing concept in E.ON Kernkraft and its sound application to ageing management issues by focussing on group 2 components such as safety relevant valves. It is demonstrated how the maintenance concept of safety relevant valves is supported by a valve diagnostic system accompanied by an applied procedure to assess the measured results with respect to the required functionality and ageing phenomena. Furthermore this concept has been developed to optimize the existing preventive maintenance of the safety relevant valves by implementing condition oriented aspects derived from the diagnostic results. The main issue of this maintenance concept is to demonstrate the high level of the secured function, reliability and performance of the safety relevant valves within an integrated ageing management. Additionally it offers improvements of all preventive maintenance issues as maintenance periods and the component related volume, spare parts management and costs. (author)

  6. Safety features in nuclear power plants to eliminate the need of ...

    Indian Academy of Sciences (India)

    on land-use were implemented and controls of radioactive contamination in foodstuffs and ... Acute Radiation Syndrome (ARS) and two died from industrial injuries. ... cerning the safety of crops, milk, food, and water; the effects of radiation ...

  7. 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

  8. Safety culture of complex risky systems: the Nuclear Engineering Institute case study

    International Nuclear Information System (INIS)

    Obadia, Isaac Jose; Vidal, Mario Cesar Rodriguez; Melo, Paulo Fernando F. Frutuoso e

    2002-01-01

    Analysis of industrial accidents have demonstrated that safe and reliable operation of complex industrial processes that use risky technology and/or hazard material depends not only on technical factors but on human and organizational factors as well. After the Chernobyl nuclear accident in 1986, the International Atomic Energy Agency established the safety culture concept and started a safety culture enhancement program within nuclear organizations worldwide. The Nuclear Engineering Institute, IEN, is a research and technological development unit of the Brazilian Nuclear Energy Commission, CNEN, characterized as a nuclear and radioactive installation where processes presenting risks to operators and to the environment are executed. In 1999, IEN started a management change program, aiming to achieve excellence of performance, based on the Model of Excellence of the National Quality Award. IEN's safety culture project is based on IAEA methodology and has been incorporated to the organizational management process. This work presents IEN's safety culture project; the results obtained on the initial safety culture assessment and the following project actions. (author)

  9. Design and construction of safety devices utilizing methods of measurement and control engineering

    Energy Technology Data Exchange (ETDEWEB)

    Greiner, B; Weidlich, S

    1982-08-01

    This article considers a proposed concept for the design and construction of measurement and control devices for the safety of chemical plants with the aim of preventing danger to persons and the environment and damage. Such measurement and control devices are generally employed when primary measures adopted for plant safety, such as safety valves, collection vessels, etc. are not applicable or insufficient by themselves. The concept regards the new sheet no. 3 of the VDI/VDE code draft 2180 ''Safety of chemical engineering plant'' and proposes a further subdivision of class A into safety classes A0, A1, and A2. Overall, it is possible, on the basis of the measures for raising the availability of measurement and control equipment which are presented in this article, to make selection appropriate to the potential danger involved. The proposed procedure should not, however, be regarded as a rigid scheme but rather as leading to a systematic view and supporting decisions resting on sound operating experience.

  10. North American Engineering, Procurement, Fabrication and Construction Worker Safety Climate Perception Affected by Job Position

    Directory of Open Access Journals (Sweden)

    Clint Pinion

    2018-04-01

    Full Text Available Understanding and implementing the results of Safety Climate surveys can assist in decreasing occupational injuries and illnesses. The following article presents findings of a cross-sectional study that assessed the relationship between safety climate perceptions and job position among engineering, procurement, fabrication and construction (EPFC employees using a 15-item survey. Descriptive statistics (means and frequencies and an ANACOVA (analysis of covariance were performed on a saturated model. The study had a 62% response rate. Results indicate a statistically significant in mean safety climate scores between job position among EPFC employees when controlling for years in industry and location type (i.e., construction versus fabrication [F (9, 603 = 5.28, p < 0.0001, adjusted R-square = 0.07]. Employee perception of safety climate differed based on the employee’s job position (i.e., laborer, foreman, etc.. Project management reported the highest safety climate scores (0.91, followed by supervisors (0.86, technical support employees and foremen (0.84 and laborers (0.81.

  11. On the development of an International Curriculum on Hydrogen Safety Engineering and its Implementation into Educational Programmes

    International Nuclear Information System (INIS)

    Dahoe, A.E.; Molkov, V.V.

    2006-01-01

    The present paper provides an overview of the development of an International Curriculum on Hydrogen Safety Engineering and its implementation into new educational programmes. The curriculum has a modular structure, and consists of five basic, six fundamental and four applied modules. The reasons for this particular structure are explained. To accelerate the development of teaching materials and their implementation in training/educational programmes, an annual European Summer School on Hydrogen Safety will be held (the first Summer School is from 15-24 Aug 2006, Belfast, UK), where leading experts deliver keynote lectures to an audience of researchers on topics covering the state-of-the-art in Hydrogen Safety Science and Engineering. The establishment of a Postgraduate Certificate course in Hydrogen Safety Engineering at the University of Ulster (starting in September 2006) as a first step in the development of a worldwide system of Hydrogen Safety education and training is described. (authors)

  12. Distinguishing feature of metal oxide films' structural engineering for gas sensor applications

    International Nuclear Information System (INIS)

    Korotcenkov, G; Golovanov, V; Brinzari, V; Cornet, A; Morante, J; Ivanov, M

    2005-01-01

    The different methods of structural engineering, used for improvement of solid state gas sensors parameters are reviewed in this paper. The wide possibilities of structural engineering in optimization of gas sensing properties were demonstrated on the example of thin tin dioxide films deposited by spray pyrolysis

  13. More safety for emergency diesel engines for the Belgium nuclear power plants

    International Nuclear Information System (INIS)

    Laire, Ch.; Scauflaire, O.; D'ans, G.; Moland, G. de; Bresseleers, J.

    2002-01-01

    Each nuclear plant in Belgium is equipped with a series of ultimate power supply (UPS) units, also called emergency power units. These consist of generators driven by multi-cylinder (typically 18) diesel engines, which are marine derivatives. Unlike marine applications, the steady-state load does not produce pulsating torques. However, these diesel engines are designed to start upon short notice following a blackout and reach full power within a few seconds to guarantee the availability or all safety valves and ventilators. Such sharp and quasi-cold starts, periodically performed to guarantee the UPS availability, may spell utter failures of the crank shaft, as demonstrated by a fatigue failure observed on the fillets connecting the crank pin to the web faces. The fillet cracks initiate in bending mode and then progress in torsion mode to excessive transient torques arising in the power train during successive starts. Aware of the potential risk and conforming to the Belgian nuclear safety rules, the plant operator of Doel sponsored the development of a nondestructive technique enabling the inspection of each fillet for cracks without first removing each piston rod from its crank pin. As a result, Laborelec developed a specific eddy-current probe which avoids fully dismantling the engine, as is done during ten-yearly overhauls with dye-checks for cracks. Inspecting crank shaft fillet integrity with this least obtrusive technique requires 24 hours per engine. It can thus be performed more frequently to prevent total crank shaft failures in time and monitor the engine fatigue caused following the mandatory monthly start-up tests. This promising technique may also find marine applications. Measuring the transient torque arising between the engine and the generator showed that this reached very high values shortly after starting the engine and injecting fuel at full throttle to reach full power within seconds. The pulsating torque of the 18-cylinders engine occurring 9

  14. Effects of hydrogen burning and associated engineered safety features on containment building response

    International Nuclear Information System (INIS)

    Iyer, S.S.; Deem, R.E.

    1982-01-01

    It is established that large amounts of hydrogen can be generated during degraded core events. The burning of this hydrogen can produce resulting loads which may represent a serious challenge to containment integrity. This paper presents some perspectives on hydrogen behavior during various degraded core events for a large dry containment. The analysis addresses the hydrogen transport and its subsequent diffusion once released to the containment. Since the distribution of hydrogen in the containment depends on the rate of release and various driving forces, the effects from various subsystems (i.e. fan coolers, sprays, heat structures, etc.) are examined to determine the sensitivity of each effect on the overall containment response. The sensitivity of results due to subcompartmentalization of the containment is also examined. Effects from localized hydrogen pocketing and burning will be addressed with emphasis on its relative impact on containment integrity

  15. Investigation of the impact of low cost traffic engineering measures on road safety in urban areas.

    Science.gov (United States)

    Yannis, George; Kondyli, Alexandra; Georgopoulou, Xenia

    2014-01-01

    This paper investigates the impact of low cost traffic engineering measures (LCTEMs) on the improvement of road safety in urban areas. A number of such measures were considered, such as speed humps, woonerfs, raised intersections and other traffic calming measures, which have been implemented on one-way, one-lane roads in the Municipality of Neo Psychiko in the Greater Athens Area. Data were analysed using the before-and-after safety analysis methodology with large control group. The selected control group comprised of two Municipalities in the Athens Greater Area, which present similar road network and land use characteristics with the area considered. The application of the methodology showed that the total number of crashes presented a statistically significant reduction, which can be possibly attributed to the introduction of LCTEMs. This reduction concerns passenger cars and single-vehicle crashes and is possibly due to the behavioural improvement of drivers of 25 years old or more. The results of this research are very useful for the identification of the appropriate low cost traffic engineering countermeasures for road safety problems in urban areas.

  16. Safety effects of low-cost engineering measures. An observational study in a Portuguese multilane road.

    Science.gov (United States)

    Vieira Gomes, Sandra; Cardoso, João Lourenço

    2012-09-01

    Single carriageway multilane roads are not, in general, a very safe type of road, mainly because of the high number of seriously injured victims in head-on collisions, when compared with dual carriageway multilane roads, with a median barrier. In this paper the results of a study on the effect of the application of several low cost engineering measures, aimed at road infrastructure correction and road safety improvement on a multilane road (EN6), are presented. The study was developed by the National Laboratory of Civil Engineering (LNEC) for the Portuguese Road Administration and involved a comparison of selected aspects of motorized traffic behaviour (traffic volumes and speeds) measured in several sections of EN6, as well as monitoring of road safety developments in the same road. The applied low cost engineering measures allowed a reduction of 10% in the expected annual number of personal injury accidents and a 70% decrease in the expected annual number of head-on collisions; the expected annual frequency of accidents involving killed and seriously injured persons was reduced by 26%. Copyright © 2012 Elsevier Ltd. All rights reserved.

  17. Design data and safety features of commercial nuclear power plants including cumulative index for Volumes I--VI

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1977-01-01

    Design data, safety features, and site characteristics are summarized for 12 nuclear power units in 6 power stations in the United States. Six pages of data are presented for each station, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. In addition, an aerial perspective is presented for each plant. This volume covers plants with docket numbers 50-553 through 50-569 (Phipps Bend, Black Fox, Yellow Creek, and NEP) and two earlier plants not previously reported--Hope Creek (50-354, 50-355) and WPPSS 1 and 4 (50-460, 50-513). Indexes for this volume and the five earlier volumes are presented in three forms--by docket number, by plant name, and by participating utility

  18. The Design of Transportation Equipment in Terms of Human Capabilities. The Role of Engineering Psychology in Transport Safety.

    Science.gov (United States)

    McFarland, Ross A.

    Human factors engineering is considered with regard to the design of safety factors for aviation and highway transportation equipment. Current trends and problem areas are identified for jet air transportation and for highway transportation. Suggested solutions to transportation safety problems are developed by applying the techniques of human…

  19. Effects of organizational safety practices and perceived safety climate on PPE usage, engineering controls, and adverse events involving liquid antineoplastic drugs among nurses.

    Science.gov (United States)

    DeJoy, David M; Smith, Todd D; Woldu, Henok; Dyal, Mari-Amanda; Steege, Andrea L; Boiano, James M

    2017-07-01

    Antineoplastic drugs pose risks to the healthcare workers who handle them. This fact notwithstanding, adherence to safe handling guidelines remains inconsistent and often poor. This study examined the effects of pertinent organizational safety practices and perceived safety climate on the use of personal protective equipment, engineering controls, and adverse events (spill/leak or skin contact) involving liquid antineoplastic drugs. Data for this study came from the 2011 National Institute for Occupational Safety and Health (NIOSH) Health and Safety Practices Survey of Healthcare Workers which included a sample of approximately 1,800 nurses who had administered liquid antineoplastic drugs during the past seven days. Regression modeling was used to examine predictors of personal protective equipment use, engineering controls, and adverse events involving antineoplastic drugs. Approximately 14% of nurses reported experiencing an adverse event while administering antineoplastic drugs during the previous week. Usage of recommended engineering controls and personal protective equipment was quite variable. Usage of both was better in non-profit and government settings, when workers were more familiar with safe handling guidelines, and when perceived management commitment to safety was higher. Usage was poorer in the absence of specific safety handling procedures. The odds of adverse events increased with number of antineoplastic drugs treatments and when antineoplastic drugs were administered more days of the week. The odds of such events were significantly lower when the use of engineering controls and personal protective equipment was greater and when more precautionary measures were in place. Greater levels of management commitment to safety and perceived risk were also related to lower odds of adverse events. These results point to the value of implementing a comprehensive health and safety program that utilizes available hazard controls and effectively communicates

  20. Special features of the safety concept and design requirements applied for Angra-2 and 3

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The special features and requirements which have been applied by NUCLEN (Nuclebras Engenharia S/A) for Angra 2 and Angra 3 and which depart somewhat from the KWU (Kraftwerk Union) standard plant, are presented. (E.G.) [pt

  1. Use of reliability engineering tools in safety and risk assessment of nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias, E-mail: amandaraso@hotmail.com, E-mail: vasconv@cdtn.br, E-mail: raissaomarques@gmail.com, E-mail: soaresw@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Serviço de Tecnologia de Reatores

    2017-07-01

    Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

  2. Use of reliability engineering tools in safety and risk assessment of nuclear facilities

    International Nuclear Information System (INIS)

    Raso, Amanda Laureano; Vasconcelos, Vanderley de; Marques, Raíssa Oliveira; Soares, Wellington Antonio; Mesquita, Amir Zacarias

    2017-01-01

    Safety, reliability and availability are fundamental criteria in design, construction and operation of nuclear facilities, as nuclear power plants. Deterministic and probabilistic risk assessments of such facilities are required by regulatory authorities in order to meet licensing regulations, contributing to assure safety, as well as reduce costs and environmental impacts. Probabilistic Risk Assessment has become an important part of licensing requirements of the nuclear power plants in Brazil and in the world. Risk can be defined as a qualitative and/or quantitative assessment of accident sequence frequencies (or probabilities) and their consequences. Risk management is a systematic application of management policies, procedures and practices to identify, analyze, plan, implement, control, communicate and document risks. Several tools and computer codes must be combined, in order to estimate both probabilities and consequences of accidents. Event Tree Analysis (ETA), Fault Tree Analysis (FTA), Reliability Block Diagrams (RBD), and Markov models are examples of evaluation tools that can support the safety and risk assessment for analyzing process systems, identifying potential accidents, and estimating consequences. Because of complexity of such analyzes, specialized computer codes are required, such as the reliability engineering software develop by Reliasoft® Corporation. BlockSim (FTA, RBD and Markov models), RENO (ETA and consequence assessment), Weibull++ (life data and uncertainty analysis), and Xfmea (qualitative risk assessment) are some codes that can be highlighted. This work describes an integrated approach using these tools and software to carry out reliability, safety, and risk assessment of nuclear facilities, as well as, and application example. (author)

  3. The arrangement of deformation monitoring project and analysis of monitoring data of a hydropower engineering safety monitoring system

    Science.gov (United States)

    Wang, Wanshun; Chen, Zhuo; Li, Xiuwen

    2018-03-01

    The safety monitoring is very important in the operation and management of water resources and hydropower projects. It is the important means to understand the dam running status, to ensure the dam safety, to safeguard people’s life and property security, and to make full use of engineering benefits. This paper introduces the arrangement of engineering safety monitoring system based on the example of a water resource control project. The monitoring results of each monitoring project are analyzed intensively to show the operating status of the monitoring system and to provide useful reference for similar projects.

  4. Quantifying the Metrics That Characterize Safety Culture of Three Engineered Systems

    International Nuclear Information System (INIS)

    Tucker, Julie; Ernesti, Mary; Tokuhiro, Akira

    2002-01-01

    With potential energy shortages and increasing electricity demand, the nuclear energy option is being reconsidered in the United States. Public opinion will have a considerable voice in policy decisions that will 'road-map' the future of nuclear energy in this country. This report is an extension of the last author's work on the 'safety culture' associated with three engineered systems (automobiles, commercial airplanes, and nuclear power plants) in Japan and the United States. Safety culture, in brief is defined as a specifically developed culture based on societal and individual interpretations of the balance of real, perceived, and imagined risks versus the benefits drawn from utilizing a given engineered systems. The method of analysis is a modified scale analysis, with two fundamental Eigen-metrics, time- (t) and number-scales (N) that describe both engineered systems and human factors. The scale analysis approach is appropriate because human perception of risk, perception of benefit and level of (technological) acceptance are inherently subjective, therefore 'fuzzy' and rarely quantifiable in exact magnitude. Perception of risk, expressed in terms of the psychometric factors 'dread risk' and 'unknown risk', contains both time- and number-scale elements. Various engineering system accidents with fatalities, reported by mass media are characterized by t and N, and are presented in this work using the scale analysis method. We contend that level of acceptance infers a perception of benefit at least two orders larger magnitude than perception of risk. The 'amplification' influence of mass media is also deduced as being 100- to 1000-fold the actual number of fatalities/serious injuries in a nuclear-related accident. (authors)

  5. Application and problems of probability methods in technical safety assessment in the field of nuclear engineering and other technologies

    International Nuclear Information System (INIS)

    Heuser, F.W.

    1980-01-01

    On the basis of a deterministic safety concept that has been developed in nuclear engineering, approaches for a probabilistic interpretation of existing safety requirements and for a further risk assessment are described. The procedures in technical reliability analysis and its application in nuclear engineering are discussed. By the example of a reliability analysis for a reactor protection system the author discusses the question as to what extent methods of reliability analysis can be used to interpret deterministically derived safety requirements. The the author gives a survey of the current value and application of probabilistic reliability assessments in non-nuclear technology. The last part of this report deals with methods of risk analysis and its use for safety assessment in nuclear engineering. On the basis of WASH 1,400 the most important phases and tasks of research work in risk assessment are explained, showing the basic criteria and the methods to be applied in risk analysis. (orig./HSCH) [de

  6. Data-driven technology for engineering systems health management design approach, feature construction, fault diagnosis, prognosis, fusion and decisions

    CERN Document Server

    Niu, Gang

    2017-01-01

    This book introduces condition-based maintenance (CBM)/data-driven prognostics and health management (PHM) in detail, first explaining the PHM design approach from a systems engineering perspective, then summarizing and elaborating on the data-driven methodology for feature construction, as well as feature-based fault diagnosis and prognosis. The book includes a wealth of illustrations and tables to help explain the algorithms, as well as practical examples showing how to use this tool to solve situations for which analytic solutions are poorly suited. It equips readers to apply the concepts discussed in order to analyze and solve a variety of problems in PHM system design, feature construction, fault diagnosis and prognosis.

  7. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Sevy, R.H.; Su, S.F.

    1985-01-01

    This paper presents the results of a study of the effectivness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). Results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs

  8. Safety design features for current UK advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-01-15

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

  9. Safety design features for current UK advanced gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  10. Features, events, processes, and safety factor analysis applied to a near-surface low-level radioactive waste disposal facility

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, M.E.; Dolinar, G.M.; Lange, B.A. [Atomic Energy of Canada Limited, Ontario (Canada)] [and others

    1995-12-31

    An analysis of features, events, processes (FEPs) and other safety factors was applied to AECL`s proposed IRUS (Intrusion Resistant Underground Structure) near-surface LLRW disposal facility. The FEP analysis process which had been developed for and applied to high-level and transuranic disposal concepts was adapted for application to a low-level facility for which significant efforts in developing a safety case had already been made. The starting point for this process was a series of meetings of the project team to identify and briefly describe FEPs or safety factors which they thought should be considered. At this early stage participants were specifically asked not to screen ideas. This initial list was supplemented by selecting FEPs documented in other programs and comments received from an initial regulatory review. The entire list was then sorted by topic and common issues were grouped, and issues were classified in three priority categories and assigned to individuals for resolution. In this paper, the issue identification and resolution process will be described, from the initial description of an issue to its resolution and inclusion in the various levels of the safety case documentation.

  11. Safety issues for LMFBR: important features drawn from the assessments of Superphenix

    International Nuclear Information System (INIS)

    Natta, M.

    2002-01-01

    Superphenix, which is built on the site of Creys-Malville, is still the biggest LMFBR plant that has been in operation. It is a pool type reactor, as Phenix and the RNR 1 500 and EFR projects. After the analysis of the preliminary safety (1974-1975), the construction was authorised by decree of the Prime Minister in 1977, the authorization for fuel loading and star-up to 3% was given by the minister of industry in July 1985 and full power was achieved in December 1986. The plant was operated until the end of December 1996, producing the equivalent of 320 EFPD, corresponding to half of the maximum barn-up of the first core. The plant was definitively stopped on the 20. of April 1998 by a decision of the French government. During this period of 25 years of licensing, construction and operation of Superphenix, others discussions and preliminary licensing procedures were started for new projects, mainly the RNR 1500 French project and the EFR European project. The operation of Superphenix was also marked by several incidents, which led to additional licensing procedures and important modifications. This period was also marked by an important work of research and development in the safety field, mostly related to the issues concerning hypothetical core disruptive accidents (HCDA) and sodium fires; further, this period was marked by the Three Mile Island accident in 1979 and the Chernobyl accident in 1986. The purpose of this paper is to present some items which were discussed during this period of 25 years and which should be of interest for future LMFBRs. In this presentation, we shall discuss the key issues concerning the safety criteria and options taken with respect to severe accidents, i.e. core melt accidents, giving details on some specific which are less known since they were assessed only lately for Superphenix, sometimes in connection with the on-going safety researches. (author)

  12. 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

  13. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

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

  14. GENIUS/EnginFrame Grid Portal: VOMS Proxy creation, new features and enhancements

    International Nuclear Information System (INIS)

    Ardizzone, V.; Barbaram, R.; Falzone, A.; Emidio, G.; Neri, L.; Scardaci, D.; Venuti, N.

    2007-01-01

    Scientific domain knowledge and tools must be presented to the (non-expert) users in terms of applications without needing to know the underlying details of the Grid Middle wares. GENIUS Grid portal, powered by EnginFrame, is an increasingly popular mechanism for creating customisable, Web-based interfaces to Grid services and resources. This work describes new GENIUS portals capabilities such as portal login, access control, display management, new approach to building reusable portal components as plug ins, better performance results as consequence of EnginFrame core functions improvement. Finally will be described two new tools developed for VOMS Proxy creation and simple JDL composer. (Author)

  15. GENIUS/EnginFrame Grid Portal: VOMS Proxy creation, new features and enhancements

    Energy Technology Data Exchange (ETDEWEB)

    Ardizzone, V.; Barbaram, R.; Falzone, A.; Emidio, G.; Neri, L.; Scardaci, D.; Venuti, N.

    2007-07-01

    Scientific domain knowledge and tools must be presented to the (non-expert) users in terms of applications without needing to know the underlying details of the Grid Middle wares. GENIUS Grid portal, powered by EnginFrame, is an increasingly popular mechanism for creating customisable, Web-based interfaces to Grid services and resources. This work describes new GENIUS portals capabilities such as portal login, access control, display management, new approach to building reusable portal components as plug ins, better performance results as consequence of EnginFrame core functions improvement. Finally will be described two new tools developed for VOMS Proxy creation and simple JDL composer. (Author)

  16. HTGR safety philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Joksimovic, V.; Fisher, C. R. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.

  17. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joksimovic, V.; Fisher, C.R.

    1981-01-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity. (author)

  18. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joskimovic, V.; Fisher, C.R.

    1980-08-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity

  19. Dynamic modeling of the tradeoff between productivity and safety in critical engineering systems

    International Nuclear Information System (INIS)

    Cowing, Michelle M.; Elisabeth Pate-Cornell, M.; Glynn, Peter W.

    2004-01-01

    Short-term tradeoffs between productivity and safety often exist in the operation of critical facilities such as nuclear power plants, offshore oil platforms, or simply individual cars. For example, interruption of operations for maintenance on demand can decrease short-term productivity but may be needed to ensure safety. Operations are interrupted for several reasons: scheduled maintenance, maintenance on demand, response to warnings, subsystem failure, or a catastrophic accident. The choice of operational procedures (e.g. timing and extent of scheduled maintenance) generally affects the probabilities of both production interruptions and catastrophic failures. In this paper, we present and illustrate a dynamic probabilistic model designed to describe the long-term evolution of such a system through the different phases of operation, shutdown, and possibly accident. The model's parameters represent explicitly the effects of different components' performance on the system's safety and reliability through an engineering probabilistic risk assessment (PRA). In addition to PRA, a Markov model is used to track the evolution of the system and its components through different performance phases. The model parameters are then linked to different operations strategies, to allow computation of the effects of each management strategy on the system's long-term productivity and safety. Decision analysis is then used to support the management of the short-term trade-offs between productivity and safety in order to maximize long-term performance. The value function is that of plant managers, within the constraints set by local utility commissions and national (e.g. energy) agencies. This model is illustrated by the case of outages (planned and unplanned) in nuclear power plants to show how it can be used to guide policy decisions regarding outage frequency and plant lifetime, and more specifically, the choice of a reactor tripping policy as a function of the state of the

  20. [Examination of safety improvement by failure record analysis that uses reliability engineering].

    Science.gov (United States)

    Kato, Kyoichi; Sato, Hisaya; Abe, Yoshihisa; Ishimori, Yoshiyuki; Hirano, Hiroshi; Higashimura, Kyoji; Amauchi, Hiroshi; Yanakita, Takashi; Kikuchi, Kei; Nakazawa, Yasuo

    2010-08-20

    How the maintenance checks of the medical treatment system, including start of work check and the ending check, was effective for preventive maintenance and the safety improvement was verified. In this research, date on the failure of devices in multiple facilities was collected, and the data of the trouble repair record was analyzed by the technique of reliability engineering. An analysis of data on the system (8 general systems, 6 Angio systems, 11 CT systems, 8 MRI systems, 8 RI systems, and the radiation therapy system 9) used in eight hospitals was performed. The data collection period assumed nine months from April to December 2008. Seven items were analyzed. (1) Mean time between failures (MTBF) (2) Mean time to repair (MTTR) (3) Mean down time (MDT) (4) Number found by check in morning (5) Failure generation time according to modality. The classification of the breakdowns per device, the incidence, and the tendency could be understood by introducing reliability engineering. Analysis, evaluation, and feedback on the failure generation history are useful to keep downtime to a minimum and to ensure safety.

  1. Engineering Technology Showcase to feature high-tech products from 18 companies

    OpenAIRE

    Gilbert, Karen

    2007-01-01

    The Student Technology Council (STC) at Virginia Tech is sponsoring an Engineering Technology Showcase on Tuesday, March 27. In addition to providing a platform for technology companies to show off their most recent innovations, technology presentations will be offered by Virginia Tech faculty and staff on topics ranging from a virtual greenhouse to the System X supercomputer.

  2. Distributed Propulsion featuring Boundary Layer Ingestion Engines for the Blended Wing Body Subsonic Transport

    NARCIS (Netherlands)

    Kok, H.J.M.; Voskuijl, M.; Van Tooren, M.J.L.

    2010-01-01

    The blended wing body aircraft is one of the promising contenders for the next generation large transport aircraft. This aircraft is particularly suitable for the use of boundary layer ingestion engines. Results published in literature suggest that it might be beneficial to have a large number of

  3. Implementing Motivational Features in Reactive Blended Learning: Application to an Introductory Control Engineering Course

    Science.gov (United States)

    Mendez, J. A.; Gonzalez, E. J.

    2011-01-01

    This paper presents a significant advance in a reactive blended learning methodology applied to an introductory control engineering course. This proposal was based on the inclusion of a reactive element (a fuzzy-logic-based controller) designed to regulate the workload for each student according to his/her activity and performance. The…

  4. Vaccine adverse event text mining system for extracting features from vaccine safety reports.

    Science.gov (United States)

    Botsis, Taxiarchis; Buttolph, Thomas; Nguyen, Michael D; Winiecki, Scott; Woo, Emily Jane; Ball, Robert

    2012-01-01

    To develop and evaluate a text mining system for extracting key clinical features from vaccine adverse event reporting system (VAERS) narratives to aid in the automated review of adverse event reports. Based upon clinical significance to VAERS reviewing physicians, we defined the primary (diagnosis and cause of death) and secondary features (eg, symptoms) for extraction. We built a novel vaccine adverse event text mining (VaeTM) system based on a semantic text mining strategy. The performance of VaeTM was evaluated using a total of 300 VAERS reports in three sequential evaluations of 100 reports each. Moreover, we evaluated the VaeTM contribution to case classification; an information retrieval-based approach was used for the identification of anaphylaxis cases in a set of reports and was compared with two other methods: a dedicated text classifier and an online tool. The performance metrics of VaeTM were text mining metrics: recall, precision and F-measure. We also conducted a qualitative difference analysis and calculated sensitivity and specificity for classification of anaphylaxis cases based on the above three approaches. VaeTM performed best in extracting diagnosis, second level diagnosis, drug, vaccine, and lot number features (lenient F-measure in the third evaluation: 0.897, 0.817, 0.858, 0.874, and 0.914, respectively). In terms of case classification, high sensitivity was achieved (83.1%); this was equal and better compared to the text classifier (83.1%) and the online tool (40.7%), respectively. Our VaeTM implementation of a semantic text mining strategy shows promise in providing accurate and efficient extraction of key features from VAERS narratives.

  5. Failure Modes Effects and Criticality Analysis, an Underutilized Safety, Reliability, Project Management and Systems Engineering Tool

    Science.gov (United States)

    Mullin, Daniel Richard

    2013-09-01

    The majority of space programs whether manned or unmanned for science or exploration require that a Failure Modes Effects and Criticality Analysis (FMECA) be performed as part of their safety and reliability activities. This comes as no surprise given that FMECAs have been an integral part of the reliability engineer's toolkit since the 1950s. The reasons for performing a FMECA are well known including fleshing out system single point failures, system hazards and critical components and functions. However, in the author's ten years' experience as a space systems safety and reliability engineer, findings demonstrate that the FMECA is often performed as an afterthought, simply to meet contract deliverable requirements and is often started long after the system requirements allocation and preliminary design have been completed. There are also important qualitative and quantitative components often missing which can provide useful data to all of project stakeholders. These include; probability of occurrence, probability of detection, time to effect and time to detect and, finally, the Risk Priority Number. This is unfortunate as the FMECA is a powerful system design tool that when used effectively, can help optimize system function while minimizing the risk of failure. When performed as early as possible in conjunction with writing the top level system requirements, the FMECA can provide instant feedback on the viability of the requirements while providing a valuable sanity check early in the design process. It can indicate which areas of the system will require redundancy and which areas are inherently the most risky from the onset. Based on historical and practical examples, it is this author's contention that FMECAs are an immense source of important information for all involved stakeholders in a given project and can provide several benefits including, efficient project management with respect to cost and schedule, system engineering and requirements management

  6. Payload Safety: Risk and Characteristic-Based Control of Engineered Nanomaterials

    Science.gov (United States)

    Abou, Seraphin Chally; Saad, Maarouf

    2013-09-01

    In the last decade progress has been made to assist organizations that are developing payloads intended for flight on the International Space Station (ISS) and/or Space Shuttle. Collaboration programs for comprehensive risk assessment have been initiated between the U.S. and the European Union to generate requirements and data needed to comply with payloads safety and to perform risk assessment and controls guidance. Yet, substantial research gaps remain, as do challenges in the translation of these research findings to control for exposure to nanoscale material payloads, and the health effects. Since nanomaterial structures are different from traditional molecules, some standard material properties can change at size of 50nm or less. Changes in material properties at this scale challenge our understanding of hazards posed by nanomaterial payloads in the ISS realistic exposure conditions, and our ability to anticipate, evaluate, and control potential health issues, and safety. The research question addressed in this framework is: what kind of descriptors can be developed for nanomaterial payloads risks assessment? Methods proposed incorporate elements of characteristic- based risk an alysis: (1) to enable characterization of anthropogenic nanomaterials which can result in incidental from natural nanoparticles; and (2) to better understand safety attributes in terms of human health impacts from exposure to varying types of engineered nanomaterials.

  7. Safety Research Experiment Facilities, Idaho National Engineering Laboratory, Idaho. Draft environmental statement

    International Nuclear Information System (INIS)

    1977-01-01

    This environmental statement was prepared in accordance with the National Environmental Policy Act of 1969 (NEPA) in support of the Energy Research and Development Administration's (ERDA) proposal for legislative authorization and appropriations for the Safety Research Experiment Facilities (SAREF) Project. The purpose of the proposed project is to modify some existing facilities and provide a new test facility at the Idaho National Engineering Laboratory (INEL) for conducting fast breeder reactor (FBR) safety experiments. The SAREF Project proposal has been developed after an extensive study which identified the FBR safety research needs requiring in-reactor experiments and which evaluated the capability of various existing and new facilities to meet these needs. The proposed facilities provide for the in-reactor testing of large bundles of prototypical FBR fuel elements under a wide variety of conditions, ranging from those abnormal operating conditions which might be expected to occur during the life of an FBR power plant to the extremely low probability, hypothetical accidents used in the evalution of some design options and in the assessment of the long-term potential risk associated with wide-scale deployment of the FBR

  8. Improving the efficacy and safety of engineered T cell therapy for cancer.

    Science.gov (United States)

    Shi, Huan; Liu, Lin; Wang, Zhehai

    2013-01-28

    Adoptive T-cell therapy (ACT) using tumor-infiltrating lymphocytes (TILs) is a powerful immunotherapeutics approach against metastatic melanoma. The success of TIL therapy has led to novel strategies for redirecting normal T cells to recognize tumor-associated antigens (TAAs) by genetically engineering tumor antigen-specific T cell receptors (TCRs) or chimeric antigen receptor (CAR) genes. In this manner, large numbers of antigen-specific T cells can be rapidly generated compared with the longer term expansion of TILs. Great efforts have been made to improve these approaches. Initial clinical studies have demonstrated that genetically engineered T cells can mediate tumor regression in vivo. In this review, we discuss the development of TCR and CAR gene-engineered T cells and the safety concerns surrounding the use of these T cells in patients. We highlight the importance of judicious selection of TAAs for modified T cell therapy and propose solutions for potential "on-target, off-organ" toxicity. Copyright © 2012 Elsevier Ireland Ltd. All rights reserved.

  9. A review of the safety features of 6M packagings for DOE programs

    International Nuclear Information System (INIS)

    1988-12-01

    This report, prepared by a US Department of Energy (DOE) Task Force and organized for clarity into two-page modules, argues that the US Department of Transportation (DOT) Specification-6M packagings (hereafter referred to as 6M packaging, or simply 6M) merit continued DOE use and, if necessary, DOE certification. This report is designed to address the specific requirements of a Safety Analysis Report for Packaging (SARP). While not a SARP, this report constitutes a compilation of all available documentation on 6M packagings. The authors individually, and the Task Force collectively, believe their investigation provides justification for the continued use of 6M packagings because they meet criteria for quality assurance and for safety under normal and accident conditions as defined by the US Nuclear Regulatory Commission (NRC) regulations. This report may be used by DOE managers to assist in deliberations on future requirements for 6M packagings as they are required to support DOE programs. For the purpose of ready evaluation, this report includes categorical topics found in Nuclear Regulatory Guide 7.9, the topical guideline for SARPs. The format, however, will (it is hoped) pleasantly surprise customary reader expectations. For, while maintaining categorical headings and subheadings found in SARPs as a skeleton, the Task Force chose to adopt the document design principles developed by Hughes Aircraft in the 1960s, ''Sequential Thematic Organization of Publications'' (STOP). 37 figs

  10. Test study on safety features of station blackout accident for nuclear main pump

    International Nuclear Information System (INIS)

    Liu Xiajie; Wang Dezhong; Zhang Jige; Liu Junsheng; Yang Zhe

    2009-01-01

    The theoretical and experimental studies of reactor coolant pump accidents encountered nation-wide and world-wide were described. To investigate the transient hydrodynamic performance of reactor coolant pump (RCP) during the period of rotational inertia in the station blackout accident, some theoretical and experimental studies were carried out, and the analysis of the test results was presented. The experiment parameters, conditions and test methods were introduced. The flow-rate, rotate speed and vibrations were analyzed emphatically. The quadruplicate polynomial curve equation was used to simulate the flow-rate,rotate speed along with time. The test results indicate that the flow-rate and rotator speed decrease rapidly at the very beginning of cut power and the test results accord with the regulation of safety standard. The vibrant displacement of bearing seat is intensified at the moment of lose power, but after a certain period rotor shaft libration changes. The test and analysis results help to understand the hydrodynamic performance of nuclear primary pump under lost of power accident, and provide the basic reference for safety evaluation. (authors)

  11. Feature engineering and a proposed decision-support system for systematic reviewers of medical evidence.

    Directory of Open Access Journals (Sweden)

    Tanja Bekhuis

    Full Text Available Evidence-based medicine depends on the timely synthesis of research findings. An important source of synthesized evidence resides in systematic reviews. However, a bottleneck in review production involves dual screening of citations with titles and abstracts to find eligible studies. For this research, we tested the effect of various kinds of textual information (features on performance of a machine learning classifier. Based on our findings, we propose an automated system to reduce screeing burden, as well as offer quality assurance.We built a database of citations from 5 systematic reviews that varied with respect to domain, topic, and sponsor. Consensus judgments regarding eligibility were inferred from published reports. We extracted 5 feature sets from citations: alphabetic, alphanumeric(+, indexing, features mapped to concepts in systematic reviews, and topic models. To simulate a two-person team, we divided the data into random halves. We optimized the parameters of a Bayesian classifier, then trained and tested models on alternate data halves. Overall, we conducted 50 independent tests.All tests of summary performance (mean F3 surpassed the corresponding baseline, P<0.0001. The ranks for mean F3, precision, and classification error were statistically different across feature sets averaged over reviews; P-values for Friedman's test were .045, .002, and .002, respectively. Differences in ranks for mean recall were not statistically significant. Alphanumeric(+ features were associated with best performance; mean reduction in screening burden for this feature type ranged from 88% to 98% for the second pass through citations and from 38% to 48% overall.A computer-assisted, decision support system based on our methods could substantially reduce the burden of screening citations for systematic review teams and solo reviewers. Additionally, such a system could deliver quality assurance both by confirming concordant decisions and by naming

  12. The Specific Features of design and process engineering in branch of industrial enterprise

    Science.gov (United States)

    Sosedko, V. V.; Yanishevskaya, A. G.

    2017-06-01

    Production output of industrial enterprise is organized in debugged working mechanisms at each stage of product’s life cycle from initial design documentation to product and finishing it with utilization. The topic of article is mathematical model of the system design and process engineering in branch of the industrial enterprise, statistical processing of estimated implementation results of developed mathematical model in branch, and demonstration of advantages at application at this enterprise. During the creation of model a data flow about driving of information, orders, details and modules in branch of enterprise groups of divisions were classified. Proceeding from the analysis of divisions activity, a data flow, details and documents the state graph of design and process engineering was constructed, transitions were described and coefficients are appropriated. To each condition of system of the constructed state graph the corresponding limiting state probabilities were defined, and also Kolmogorov’s equations are worked out. When integration of sets of equations of Kolmogorov the state probability of system activity the specified divisions and production as function of time in each instant is defined. On the basis of developed mathematical model of uniform system of designing and process engineering and manufacture, and a state graph by authors statistical processing the application of mathematical model results was carried out, and also advantage at application at this enterprise is shown. Researches on studying of loading services probability of branch and third-party contractors (the orders received from branch within a month) were conducted. The developed mathematical model of system design and process engineering and manufacture can be applied to definition of activity state probability of divisions and manufacture as function of time in each instant that will allow to keep account of loading of performance of work in branches of the enterprise.

  13. Adding a Visualization Feature to Web Search Engines: It’s Time

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Pak C.

    2008-11-11

    Since the first world wide web (WWW) search engine quietly entered our lives in 1994, the “information need” behind web searching has rapidly grown into a multi-billion dollar business that dominates the internet landscape, drives e-commerce traffic, propels global economy, and affects the lives of the whole human race. Today’s search engines are faster, smarter, and more powerful than those released just a few years ago. With the vast investment pouring into research and development by leading web technology providers and the intense emotion behind corporate slogans such as “win the web” or “take back the web,” I can’t help but ask why are we still using the very same “text-only” interface that was used 13 years ago to browse our search engine results pages (SERPs)? Why has the SERP interface technology lagged so far behind in the web evolution when the corresponding search technology has advanced so rapidly? In this article I explore some current SERP interface issues, suggest a simple but practical visual-based interface design approach, and argue why a visual approach can be a strong candidate for tomorrow’s SERP interface.

  14. SOME ASPECTS OF FEATURES CONCEPT AND NATURE OF ENVIRONMENTAL SAFETY LIFE URBAN POPULATION

    Directory of Open Access Journals (Sweden)

    Kirova M.

    2017-05-01

    Full Text Available Proved the concept and nature of the environmental safety of life of the urban population. A structural levels in the list of objects to enter the security levels of urban or rural communities of local communities, which changes the nature of the institutional environment environmental security. In assessing the quality of life of urban society used many parameters of socio-economic and environmental. General rules on the prevention of environmental degradation and risks to human health set forth in the applicable law "On Environmental Protection", on the basis of which developed a number of legal documents. As for the capital of Ukraine - Kyiv, it is a great cultural, historical and commercial and industrial center. So we can safely say that Kyiv is characterized by all the environmental problems that are inherent in all large cities. This, above all, traffic pollution, changes in the air quality, noise pollution, emissions of chemicals into the atmosphere, pollution, toxic waste, the problem of waste. But a special place in this region occupies a radioactive problem due to the Chernobyl nuclear power plant. Also, nearby, in the Kyiv region Kiev reservoir are with not the best environmental conditions. The most important components of ecosystems: the air of the city, urban and suburban water sources, soil city. International experts conducted a study in 215 cities around the world. Kyiv international ranking is 29 in place pollution. Thus, the problem of environmental security for the population of the city. Kyiv, as many large cities in Ukraine is quite relevant. In this regard, the article studies that the environmental safety of life of the urban population in the context of national security - a state of effective security systems (environment interconnected structural levels of the individual, the local community, society, state and international (global community of aggregate factors that endanger or threaten the very existence of which

  15. Fundamental features and main problems of nuclear power and radiological safety law

    International Nuclear Information System (INIS)

    Moser, B.

    1981-01-01

    This report deals on a general basis with the legal spheres affected by the utilisation of nuclear energy and protection from ionising radiation. Following a historical survey of the development both in the field of national legisation in Austria and internationally, the five principal legal spheres are discussed in detail. These are administrative law, liability and insurance law, criminal law, constitutional law and international law. In the foreground of discussion is administrative law, which is mainly of a preventive nature. This also comprises radiological safety law. Next in importance is liability and insurance law, which, in contrast to the former, aims at compensation for damage. Criminal law is also intended to have a preventive effect. Finally, the author discusses the peaceful use of nuclear energy in relation to the constitutional law and the international law in force. (Auth.)

  16. Shielding repair of N.S. Mutsu and related safety features

    International Nuclear Information System (INIS)

    Kishimoto, K.; Miyakoshi, J.

    1978-01-01

    The abnormal radiation level observed on the upper deck of N.S. Mutsu was caused by neutrons streaming through an annular air gap between the reactor pressure vessel and the primary shield. In order to lower this level, a modification of shielding has been planned, for which a shielding mock-up experiment was carried. The foregoing modifications brought some change to the expected behavior of the reactor plant under ship accident situations, and studies were performed to verify plant safety, such as calculations to determine containment vessel integrity and decay heat removal after sinking, and calculations supported by experiment to ascertain the structural strength of the double bottom upon stranding of the ship

  17. Autonomous safety and reliability features of the K-1 avionics system

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, G.E.; Kohrs, D.; Bailey, R.; Lai, G. [Kistler Aerospace Corp., Kirkland, WA (United States)

    2004-03-01

    Kistler Aerospace Corporation is developing the K-1, a fully reusable, two-stage-to-orbit launch vehicle. Both stages return to the launch site using parachutes and airbags. Initial flight operations will occur from Woomera, Australia. K-1 guidance is performed autonomously. Each stage of the K- 1 employs a triplex, fault tolerant avionics architecture, including three fault tolerant computers and three radiation hardened Embedded GPS/INS units with a hardware voter. The K-1 has an Integrated Vehicle Health Management (IVHM) system on each stage residing in the three vehicle computers based on similar systems in commercial aircraft. During first-stage ascent, the IVHM system performs an Instantaneous Impact Prediction (IIP) calculation 25 times per second, initiating an abort in the event the vehicle is outside a predetermined safety corridor for at least three consecutive calculations. In this event, commands are issued to terminate thrust, separate the stages, dump all propellant in the first-stage, and initiate a normal landing sequence. The second-stage flight computer calculates its ability to reach orbit along its state vector, initiating an abort sequence similar to the first stage if it cannot. On a nominal mission, following separation, the second-stage also performs calculations to assure its impact point is within a safety corridor. The K-1's guidance and control design is being tested through simulation with hardware-in-the-loop at Draper Laboratory. Kistler's verification strategy assures reliable and safe operation of the K-1. (author)

  18. Application of condition-based HRA method for a manual actuation of the safety features in a nuclear power Plant

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung-Cheol

    2006-01-01

    A practical approach to develop a more realistic fault-tree model with a consideration of various conditions endured by a human operator is proposed. In safety-critical systems, the generation failure of an actuation signal is caused by the concurrent failures of the automated systems and an operator action. These two sources of safety signals are complicatedly correlated. The failures of sensors or automated systems will cause a lack of necessary information for a human operator and result in error-forcing contexts such as the loss of corresponding alarms and indications. It is well known that the error-forcing contexts largely affect the operator's performance. An automated system which consists of multiple processing channels and complex components is also affected by the availability of the sensors. This paper proposes a condition-based human reliability assessment (CBHRA) method in order to address these complicated conditions in a practical way. We apply the CBHRA method to the manual actuation of the safety features such as a reactor trip and auxiliary feedwater actuation in Korean Standard Nuclear Power Plants. Even the human error probability of each given condition is simply assumed, the application results prove that the CBHRA effectively accommodates the complicated error-forcing contexts into the fault trees

  19. Technical and institutional safety features of nuclear power plants in Brazil. Aspectos tecnicos e institucionais da seguranca dos reatores nucleares no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, L P [Sociedade Brasileira de Fisica, Rio de Janeiro, RJ (Brazil)

    1986-01-01

    This work reports technical, political and institutional safety features of nuclear power plants in Brazil. It is mainly concerned with reactor accidents and personnel safety. The three mile Island and Chernobyl accidents are also discussed and taken as examples. (A.C.A.S.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  1. Safety analysis of dual purpose metal cask subjected to impulsive loads due to aircraft engine crash

    International Nuclear Information System (INIS)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    2009-01-01

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters and seismic tests subjected to strong earthquake motions. Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001. This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine research (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are developed

  2. Safety Analysis of Dual Purpose Metal Cask Subjected to Impulsive Loads due to Aircraft Engine Crash

    Science.gov (United States)

    Shirai, Koji; Namba, Kosuke; Saegusa, Toshiari

    In Japan, the first Interim Storage Facility of spent nuclear fuel away from reactor site is being planned to start its commercial operation around 2010, in use of dual-purpose metal cask in the northern part of Main Japan Island. Business License Examination for safety design approval has started since March, 2007. To demonstrate the more scientific and rational performance of safety regulation activities on each phase for the first license procedure, CREPEI has executed demonstration tests with full scale casks, such as drop tests onto real targets without impact limiters(1) and seismic tests subjected to strong earthquake motions(2). Moreover, it is important to develop the knowledge for the inherent security of metal casks under extreme mechanical-impact conditions, especially for increasing interest since the terrorist attacks from 11th September 2001(3)-(6). This paper presents dynamic mechanical behavior of the metal cask lid closure system caused by direct aircraft engine crash and describes calculated results (especially, leak tightness based on relative dynamic displacements between metallic seals). Firstly, the local penetration damage of the interim storage facility building by a big passenger aircraft engine crash (diameter 2.7m, length 4.3m, weight 4.4ton, impact velocity 90m/s) has been examined. The reduced velocity is calculated by the local damage formula for concrete structure with its thickness of 70cm. The load vs. time function for this reduced velocity (60m/s) is estimated by the impact analysis using Finite Element code LS-DYNA with the full scale engine model onto a hypothetically rigid target. Secondly, as the most critical scenarios for the metal cask, two impact scenarios (horizontal impact hitting the cask and vertical impact onto the lid metallic seal system) are chosen. To consider the geometry of all bolts for two lids, the gasket reaction forces and the inner pressure of the cask cavity, the detailed three dimensional FEM models are

  3. A reverse engineering methodology for nickel alloy turbine blades with internal features

    DEFF Research Database (Denmark)

    Gameros, A.; De Chiffre, Leonardo; Siller, H.R.

    2015-01-01

    The scope of this work is to present a reverse engineering (RE) methodology for freeform surfaces, based on a case study of a turbine blade made of Inconel, including the reconstruction of its internal cooling system. The methodology uses an optical scanner and X-ray computed tomography (CT......) equipment. Traceability of the measurements was obtained through the use of a Modular Freeform Gage (MFG). An uncertainty budget is presented for both measuring technologies and results show that the RE methodology presented is promising when comparing uncertainty values against common industrial tolerances....

  4. Design of concrete structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety standard for civil engineering structures important to safety of nuclear facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design of concrete structures important to safety

  5. Bladder tissue engineering using biocompatible nanofibrous electrospun constructs: feasibility and safety investigation.

    Science.gov (United States)

    Shakhssalim, Nasser; Dehghan, Mohammad Mehdi; Moghadasali, Reza; Soltani, Mohammad Hossein; Shabani, Iman; Soleimani, Masoud

    2012-01-01

    To investigate the feasibility and safety of using biocompatible, nanofibrous electrospun polycaprolactone (PCL) and combination of polylactic acid (PLLA) and PCL mats in a canine model. Plasma-treated electrospun unseeded mats were implanted in three dogs. The first dog was sacrificed after 3 months and the second and third ones after 4 months, and then, the graft was examined macroscopically with subsequent morphological and histochemical evaluation. Both films showed high levels of cell infiltration and tissue formation, but body response to PLLA/PCL mat in comparison to PCL mat was very low. All three implantation models showed the same light microscopic morphology, immunohistochemistry, and scanning electron microscopy results; nevertheless, only the PCL/PLLA model showed favorable clinical results. Based on these data, nanofibrous PLLA/PCL scaffolding could be a suitable material for the bladder tissue engineering; however, it deserves further investigations.

  6. 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

  7. Safety and environmental protection - realization efforts regarding the authorization of power engineering plants

    International Nuclear Information System (INIS)

    Dreyhaupt, F.J.

    1982-01-01

    As to power plants the author calls special attention to three statements concerning questions on safety and environmental protection which play a decisive part in authorizing conventional and nuclear facilities. After investigating the most important legal fundamental principles for the authorization of power engineering plants, the Atomic Act and the Federal Immission Protection Act, the author discusses the problems that arise with the application of the authorization procedures. The reasons which can be made responsible for the long running periods of the authorization procedures and therewith of the realization of the site installation work are given. Finally, the author describes and judges the outlines of regulations for large scale furnaces and for the TA-air supplementary clause. (orig.) [de

  8. Design for safety: A cognitive engineering approach to the control and management of nuclear power plants

    International Nuclear Information System (INIS)

    Boy, Guy A.; Schmitt, Kara A.

    2013-01-01

    Highlights: ► Complexity must be understood and handled well in order to design for safety. ► Complexity can be reduced during design by using the AUTOS pyramid model. ► Procedures are human automation, much as software is machine automation. ► Identifying emergent behaviors reduces procedure accumulation. ► Human-in-the-loop-simulations help to understand emergent behaviors. -- Abstract: This paper presents an analytical approach to design for safety that is based on 30 years of experience in the field of Human-centered design. This field is often qualified as governing safety–critical systems where risk management is a crucial issue. We need to better understand what the main facets of safety are that should be taken into account during the design and development processes. There are many factors that contribute to design for safety. We propose some of these factors and an articulation of them from requirement gathering and synthesis to formative evaluations to summative evaluations. Among these factors, we analyze complexity, flexibility, stability, redundancy, support, training, experience and testing. However, we cannot design a safe and reliable product in one shot; design is incremental. A product and its various uses become progressively mature. When we deal with new products, issues come from the fact that practice features emerge from the use of the product and are difficult, even impossible, to predict ahead of time. The automation within is an important portion of this maturity, and must be understood well. This is why design for safety is not possible without anticipatory simulations and a period of tests in the real world, such as operational testing in nuclear power plants. In addition, designing for safety is not finished when the product is delivered; experience feedback, or human-in-the-loop simulation (HITLS) is an important part of the overall global design process. The AUTOS pyramid approach can assist in simplifying the

  9. An image-based method to measure all-terrain vehicle dimensions for engineering safety purposes.

    Science.gov (United States)

    Jennissen, Charles A; Miller, Nathan S; Tang, Kaiyang; Denning, Gerene M

    2014-04-01

    All-terrain vehicle (ATV) crashes are a serious public health and safety concern. Engineering approaches that address ATV injury prevention are critically needed. Avenues to pursue include evidence-based seat design that decreases risky behaviours, such as carrying passengers and operation of adult-size vehicles by children. The goal of this study was to create and validate an image-based method to measure ATV seat length and placement. Publicly available ATV images were downloaded. Adobe Photoshop was then used to generate a vertical grid through the centre of the vehicle, to define the grid scale using the manufacturer's reported wheelbase, and to determine seat length and placement relative to the front and rear axles using this scale. Images that yielded a difference greater than 5% between the calculated and the manufacturer's reported ATV lengths were excluded from further analysis. For the 77 images that met inclusion criteria, the mean±SD for the difference in calculated versus reported vehicle length was 1.8%±1.2%. The Pearson correlation coefficient for comparing image-based seat lengths determined by two independent measurers (20 models) and image-based lengths versus lengths measured at dealerships (12 models) were 0.95 and 0.96, respectively. The image-based method provides accurate and reproducible results for determining ATV measurements, including seat length and placement. This method greatly expands the number of ATV models that can be studied, and may be generalisable to other motor vehicle types. These measurements can be used to guide engineering approaches that improve ATV safety design.

  10. Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants

    International Nuclear Information System (INIS)

    O'Hara, J.M.; Higgins, J.; Fleger, Stephen

    2011-01-01

    The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.

  11. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  12. Model checking of safety-critical software in the nuclear engineering domain

    International Nuclear Information System (INIS)

    Lahtinen, J.; Valkonen, J.; Björkman, K.; Frits, J.; Niemelä, I.; Heljanko, K.

    2012-01-01

    Instrumentation and control (I and C) systems play a vital role in the operation of safety-critical processes. Digital programmable logic controllers (PLC) enable sophisticated control tasks which sets high requirements for system validation and verification methods. Testing and simulation have an important role in the overall verification of a system but are not suitable for comprehensive evaluation because only a limited number of system behaviors can be analyzed due to time limitations. Testing is also performed too late in the development lifecycle and thus the correction of design errors is expensive. This paper discusses the role of formal methods in software development in the area of nuclear engineering. It puts forward model checking, a computer-aided formal method for verifying the correctness of a system design model, as a promising approach to system verification. The main contribution of the paper is the development of systematic methodology for modeling safety critical systems in the nuclear domain. Two case studies are reviewed, in which we have found errors that were previously not detected. We also discuss the actions that should be taken in order to increase confidence in the model checking process.

  13. The future of nuclear energy. Safety and nuclear power plants. Contribution of engineering companies

    International Nuclear Information System (INIS)

    Garcia Rodriguez, A.

    1995-01-01

    Risk, its consideration and its acceptance or rejection, are parameters which to a large extent are independent and sometimes difficult to interrelate. Nuclear energy, unlike motoring or civil aviation, has not gained sufficient public acceptance, this despite the fact that the risk to population is by far the least of the three. It is therefore necessary to continue with its improvement is an attempt to create the same confidence in the nuclear industry, as society has placed in civil aviation. Improvement in future nuclear power plants must be a combination of improved safety and a reduction in capital investment. This objective can only be reached through standardization and international cooperation. Engineering has a very important part to play in the standardization process. An increase in engineering input during the design, construction, operation and maintenance phases of future nuclear power plants, and the application of increasingly sophisticated analysis and management tools are anticipated. Nevertheless, the financial impact on the cost of each plant will fall as a result of increased input spread throughout the range of standard plants. Our current Advanced Reactor Power Plant Programme enables the Spanish industry to participate actively in the creation of future standards nuclear power plants. Having a presence in selected engineering activities, which guarantees access to the state of the art in this area, is one of our priorities, since it will facilitate the presence of the rest of the industry in future projects. If the objectives described above are to be reached, the present involvement of the spanish industry in this programme must be maintained in the medium and long term. (Author)

  14. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  15. New small HTGR power plant concept with inherently safe features - an engineering and economic challenge

    International Nuclear Information System (INIS)

    McDonald, C.F.; Sonn, D.L.

    1983-01-01

    Studies are in a very early design stage to establish a modular concept High-Temperature Gas-Cooled Reactor (HTGR) plant of about 100-MW(e) size to meet the special needs of small energy users in the industrialized and developing nations. The basic approach is to design a small system in which, even under the extreme conditions of loss of reactor pressure and loss of forced core cooling, the temperature would remain low enough so that the fuel would retain essentially all the fission products and the owner's investment would not be jeopardized. To realize economic goals, the designer faces the challenge of providing a standardized nuclear heat source, relying on a high percentage of factory fabrication to reduce site construction time, and keeping the system simple. While the proposed nuclear plant concept embodies new features, there is a large technology base to draw upon for the design of a small HTGR

  16. 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.

  17. 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.

  18. Biosynthesis of ilamycins featuring unusual building blocks and engineered production of enhanced anti-tuberculosis agents.

    Science.gov (United States)

    Ma, Junying; Huang, Hongbo; Xie, Yunchang; Liu, Zhiyong; Zhao, Jin; Zhang, Chunyan; Jia, Yanxi; Zhang, Yun; Zhang, Hua; Zhang, Tianyu; Ju, Jianhua

    2017-08-30

    Tuberculosis remains one of the world's deadliest communicable diseases, novel anti-tuberculosis agents are urgently needed due to severe drug resistance and the co-epidemic of tuberculosis/human immunodeficiency virus. Here, we show the isolation of six anti-mycobacterial ilamycin congeners (1-6) bearing rare L-3-nitro-tyrosine and L-2-amino-4-hexenoic acid structural units from the deep sea-derived Streptomyces atratus SCSIO ZH16. The biosynthesis of the rare L-3-nitrotyrosine and L-2-amino-4-hexenoic acid units as well as three pre-tailoring and two post-tailoring steps are probed in the ilamycin biosynthetic machinery through a series of gene inactivation, precursor chemical complementation, isotope-labeled precursor feeding experiments, as well as structural elucidation of three intermediates (6-8) from the respective mutants. Most impressively, ilamycins E 1 /E 2 , which are produced in high titers by a genetically engineered mutant strain, show very potent anti-tuberculosis activity with an minimum inhibitory concentration value ≈9.8 nM to Mycobacterium tuberculosis H37Rv constituting extremely potent and exciting anti-tuberculosis drug leads.Tuberculosis (TB) remains one of the world's deadliest communicable diseases, novel anti-TB agents are urgently needed due to severe drug resistance and the co-epidemic of TB/HIV. Here, the authors show that anti-mycobacterial ilamycin congeners bearing unusual structural units possess extremely potent anti-tuberculosis activities.

  19. Engineering task plan for the annual revision of the rotary mode core sampling system safety equipment list

    International Nuclear Information System (INIS)

    BOGER, R.M.

    1999-01-01

    This Engineering Task Plan addresses an effort to provide an update to the RMCS Systems 3 and 4 SEL and DCM in order to incorporate the changes to the authorization basis implemented by HNF-SD-WM-BIO-001, Rev. 0 (Draft), Addendum 5 , Safety Analysis for Rotary Mode Core Sampling. Responsibilities, task description, cost estimate, and schedule are presented

  20. Exploring Barriers to Medication Safety in an Ethiopian Hospital Emergency Department: A Human Factors Engineering Approach

    Directory of Open Access Journals (Sweden)

    Ephrem Abebe

    2018-02-01

    Full Text Available Objective: To describe challenges associated with the medication use process and potential medication safety hazards in an Ethiopian hospital emergency department using a human factors approach. Methods: We conducted a qualitative study employing observations and semi-structured interviews guided by the Systems Engineering Initiative for Patient Safety model of work system as an analytical framework. The study was conducted in the emergency department of a teaching hospital in Ethiopia. Study participants included resident doctors, nurses, and pharmacists. We performed content analysis of the qualitative data using accepted procedures. Results: Organizational barriers included communication failures, limited supervision and support for junior staff contributing to role ambiguity and conflict. Compliance with documentation policy was minimal. Task related barriers included frequent interruptions and work-related stress resulting from job requirements to continuously prioritize the needs of large numbers of patients and family members. Person related barriers included limited training and work experience. Work-related fatigue due to long working hours interfered with staff’s ability to document and review medication orders. Equipment breakdowns were common as were non-calibrated or poorly maintained medical devices contributing to erroneous readings. Key environment related barriers included overcrowding and frequent interruption of staff’s work. Cluttering of the work space compounded the problem by impeding efforts to locate medications, medical supplies or medical charts. Conclusions: Applying a systems based approach allows a context specific understanding of medication safety hazards in EDs from low-income countries. When developing interventions to improve medication and overall patient safety, health leaders should consider the interactions of the different factors. Conflict of Interest We declare no conflicts of interest or

  1. An assessment system for the system safety engineering capability maturity model in the case of spent fuel reprocessing

    International Nuclear Information System (INIS)

    Yang Xiaohua; Liu Zhenghai; Liu Zhiming; Wan Yaping; Bai Xiaofeng

    2012-01-01

    We can improve the processing, the evaluation of capability and promote the user's trust by using system security engineering capability maturity model (SSE-CMM). SSE-CMM is the common method for organizing and implementing safety engineering, and it is a mature method for system safety engineering. Combining capability maturity model (CMM) with total quality management and statistic theory, SSE-CMM turns systems security engineering into a well-defined, mature, measurable, advanced engineering discipline. Lack of domain knowledge, the size of data, the diversity of evidences, the cumbersomeness of processes, and the complexity of matching evidences with problems are the main issues that SSE-CMM assessment has to face. To improve effectively the efficiency of assessment of spent fuel reprocessing system security engineering capability maturity model (SFR-SSE-CMM), in this paper we de- signed an intelligent assessment software based on domain ontology and that uses methods such as ontology, evidence theory, semantic web, intelligent information retrieval and intelligent auto-matching techniques. This software includes four subsystems, which are domain ontology creation and management system, evidence auto collection system, and a problem and evidence matching system. The architecture of the software is divided into five layers: a data layer, an oncology layer, a knowledge layer, a service layer arid a presentation layer. (authors)

  2. A comparison of safety culture associated with three engineered systems in Japan and the United States

    International Nuclear Information System (INIS)

    Tokuhiro, Akira

    2001-01-01

    The internationally reported nuclear criticality accident at JCO in Tokaimura, Japan has further eroded public confidence in nuclear energy, its related facilities and the (Japanese) government's ability to handle such a crisis. The JCO accident marked the sixth nuclear-related incident since 1995. The existing state of 'safety culture' is being questioned and re-evaluated at a national level. In this work the safety culture associated with engineered systems (ES) such as the automobile, commercial airplane and nuclear power plants (NPP) are evaluated based on a scale-analysis (SA), via proposition of two fundamental parameters called eigenmetrics. The identified eigenmetrics are time- (τ) and number-scales (N) describing both ES and human factors, at the individual and/or societal levels. The SA approach is appropriate because human perception of risk (POR), perception of benefit (POB) and level of (technology) acceptance (LOA) are inherently subjective, therefore 'fuzzy' and rarely quantifiable in exact magnitude. POR expressed in terms of the psychometric factors 'dread risk' and 'unknown risk', contain both time- and number-scale elements. The JCO accident, as well as auto-fatalities, commercial airline accidents and hypothetical NPP accidents are characterized in terms of τ, N and two additional derived parameters of relevance, Nτ and N/τ. We contend that LOA infers a POB at least two orders of magnitude larger than POR. The 'amplification' influence of mass-media is also deduced as being 100 to 1000 fold the actual number of fatalities/serious injuries in a nuclear-related accident. (author)

  3. Integrating valve-inspired design features into poly(ethylene glycol) hydrogel scaffolds for heart valve tissue engineering.

    Science.gov (United States)

    Zhang, Xing; Xu, Bin; Puperi, Daniel S; Yonezawa, Aline L; Wu, Yan; Tseng, Hubert; Cuchiara, Maude L; West, Jennifer L; Grande-Allen, K Jane

    2015-03-01

    The development of advanced scaffolds that recapitulate the anisotropic mechanical behavior and biological functions of the extracellular matrix in leaflets would be transformative for heart valve tissue engineering. In this study, anisotropic mechanical properties were established in poly(ethylene glycol) (PEG) hydrogels by crosslinking stripes of 3.4 kDa PEG diacrylate (PEGDA) within 20 kDa PEGDA base hydrogels using a photolithographic patterning method. Varying the stripe width and spacing resulted in a tensile elastic modulus parallel to the stripes that was 4.1-6.8 times greater than that in the perpendicular direction, comparable to the degree of anisotropy between the circumferential and radial orientations in native valve leaflets. Biomimetic PEG-peptide hydrogels were prepared by tethering the cell-adhesive peptide RGDS and incorporating the collagenase-degradable peptide PQ (GGGPQG↓IWGQGK) into the polymer network. The specific amounts of RGDS and PEG-PQ within the resulting hydrogels influenced the elongation, de novo extracellular matrix deposition and hydrogel degradation behavior of encapsulated valvular interstitial cells (VICs). In addition, the morphology and activation of VICs grown atop PEG hydrogels could be modulated by controlling the concentration or micro-patterning profile of PEG-RGDS. These results are promising for the fabrication of PEG-based hydrogels using anatomically and biologically inspired scaffold design features for heart valve tissue engineering. Copyright © 2014 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

  4. Optimal design of high damping force engine mount featuring MR valve structure with both annular and radial flow paths

    International Nuclear Information System (INIS)

    Nguyen, Q H; Choi, S B; Lee, Y S; Han, M S

    2013-01-01

    This paper focuses on the optimal design of a compact and high damping force engine mount featuring magnetorheological fluid (MRF). In the mount, a MR valve structure with both annular and radial flows is employed to generate a high damping force. First, the configuration and working principle of the proposed MR mount is introduced. The MRF flows in the mount are then analyzed and the governing equations of the MR mount are derived based on the Bingham plastic behavior of the MRF. An optimal design of the MR mount is then performed to find the optimal structure of the MR valve to generate a maximum damping force with certain design constraints. In addition, the gap size of MRF ducts is empirically chosen considering the ‘lockup’ problem of the mount at high frequency. Performance of the optimized MR mount is then evaluated based on finite element analysis and discussions on performance results of the optimized MR mount are given. The effectiveness of the proposed MR engine mount is demonstrated via computer simulation by presenting damping force and power consumption. (paper)

  5. Optimal design of high damping force engine mount featuring MR valve structure with both annular and radial flow paths

    Science.gov (United States)

    Nguyen, Q. H.; Choi, S. B.; Lee, Y. S.; Han, M. S.

    2013-11-01

    This paper focuses on the optimal design of a compact and high damping force engine mount featuring magnetorheological fluid (MRF). In the mount, a MR valve structure with both annular and radial flows is employed to generate a high damping force. First, the configuration and working principle of the proposed MR mount is introduced. The MRF flows in the mount are then analyzed and the governing equations of the MR mount are derived based on the Bingham plastic behavior of the MRF. An optimal design of the MR mount is then performed to find the optimal structure of the MR valve to generate a maximum damping force with certain design constraints. In addition, the gap size of MRF ducts is empirically chosen considering the ‘lockup’ problem of the mount at high frequency. Performance of the optimized MR mount is then evaluated based on finite element analysis and discussions on performance results of the optimized MR mount are given. The effectiveness of the proposed MR engine mount is demonstrated via computer simulation by presenting damping force and power consumption.

  6. The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project

    International Nuclear Information System (INIS)

    Hopper, Calvin Mitchell

    2011-01-01

    In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT and SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT and SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National

  7. NASA Engineering Safety Center NASA Aerospace Flight Battery Systems Working Group 2007 Proactive Task Status

    Science.gov (United States)

    Manzo, Michelle A.

    2007-01-01

    In 2007, the NASA Engineering Safety Center (NESC) chartered the NASA Aerospace Flight Battery Systems Working Group to bring forth and address critical battery-related performance/manufacturing issues for NASA and the aerospace community. A suite of tasks identifying and addressing issues related to Ni-H2 and Li-ion battery chemistries was submitted and selected for implementation. The current NESC funded are: (1) Wet Life of Ni-H2 Batteries (2) Binding Procurement (3) NASA Lithium-Ion Battery Guidelines (3a) Li-Ion Performance Assessment (3b) Li-Ion Guidelines Document (3b-i) Assessment of Applicability of Pouch Cells for Aerospace Missions (3b-ii) High Voltage Risk Assessment (3b-iii) Safe Charge Rates for Li-Ion Cells (4) Availability of Source Material for Li-Ion Cells (5) NASA Aerospace Battery Workshop This presentation provides a brief overview of the tasks in the 2007 plan and serves as an introduction to more detailed discussions on each of the specific tasks.

  8. Environmental, health, and safety effects of engineered nanomaterials: challenges and research needs

    Science.gov (United States)

    Fairbrother, Howard

    2010-04-01

    The number of technologies and consumer products that incorporate engineered nanomaterials (ENMs) has grown rapidly. Indeed, ENMs such as carbon nanotubes and nano-silver, are revolutionizing many commercial technologies and have already been incorporated into more than 800 commercial products, including polymer composites, cell phone batteries, sporting equipment and cosmetics. The global market for ENMs has grown steadily from 7.5 billion in 2003 to 12.7 billion in 2008. Over the next five years, their market value is expected to exceed $27 billion. This surge in demand has been responsible for a corresponding increase in the annual production rates of ENMs. For example, Bayer anticipates that single and multi-walled carbon nanotubes (SWNT and MWNT) production rates will reach 3,000 tons/yr by 2012. Inevitably, some of these synthetic materials will enter the environment either from incidental release during manufacture and transport, or following use and disposal. Consequently, intense scientific research is now being directed towards understanding the environmental, health and safety (EHS) risks posed by ENMs. I will highlight some of the key research challenges and needs in this area, include (i) developing structure-property relationships that will enable physicochemical properties of ENMs to be correlated with environmentally relevant behavior (e.g. colloidal properties, toxicity), (ii) determining the behavior of nanoproducts, and (iii) developing analytical techniques capable of detecting and quantifying the concentration of ENMs in the environment.

  9. Researches in radiation protection and safety at Moscow engineering physics institute

    International Nuclear Information System (INIS)

    Kramer-Ageev, E.A.; Lebedev, L.A.

    1994-01-01

    Department of Radiation Physics of Moscow Engineering Physics Institute is a research and teaching institution in the field of radiation protection, dosimetry, shielding and in radioecology. The scientific activity which has been doing at the department for many years includes the following directions: 1. Development of mathematical models and computational methods for an evaluation of external and internal exposure of people living on contaminated areas. Recently the computational model for forecast of internal irradiation via food chains was linked with computer geographical information systems. 2. Development of techniques and instruments for the measurements of radioactive contamination of soil, air, water and agricultural products. Department has special laboratory for this. 3. Application of computational methods to the problem of nuclear medicine. The whole body spectrometry and radiation 'coding' are used as an efficient methods of obtaining information on the radionuclides location in the human body. 4. Application of computational methods to the problem of radiation safety at nuclear power plants. It allows one to calculate radiation fields in shielding and the characteristics of nuclear wastes. (author)

  10. Engineered nanomaterials in food: implications for food safety and consumer health.

    Science.gov (United States)

    Martirosyan, Alina; Schneider, Yves-Jacques

    2014-05-28

    From the current state-of-the-art, it is clear that nanotechnology applications are expected to bring a range of benefits to the food sector aiming at providing better quality and conservation. In the meantime, a growing number of studies indicate that the exposure to certain engineered nanomaterials (ENMs) has a potential to lead to health complications and that there is a need for further investigations in order to unravel the biological outcomes of nanofood consumption. In the current review, we summarize the existing data on the (potential) use of ENMs in the food industry, information on the toxicity profiles of the commonly applied ENMs, such as metal (oxide) nanoparticles (NPs), address the potential food safety implications and health hazards connected with the consumption of nanofood. A number of health complications connected with the human exposure to ENMs are discussed, demonstrating that there is a real basis for the arisen concern not only connected with the gut health, but also with the potency to lead to systemic toxicity. The toxicological nature of hazard, exposure levels and risk to consumers from nanotechnology-derived food are on the earliest stage of investigation and this review also highlights the major gaps that need further research and regulation.

  11. Engineered Nanomaterials in Food: Implications for Food Safety and Consumer Health

    Directory of Open Access Journals (Sweden)

    Alina Martirosyan

    2014-05-01

    Full Text Available From the current state-of-the-art, it is clear that nanotechnology applications are expected to bring a range of benefits to the food sector aiming at providing better quality and conservation. In the meantime, a growing number of studies indicate that the exposure to certain engineered nanomaterials (ENMs has a potential to lead to health complications and that there is a need for further investigations in order to unravel the biological outcomes of nanofood consumption. In the current review, we summarize the existing data on the (potential use of ENMs in the food industry, information on the toxicity profiles of the commonly applied ENMs, such as metal (oxide nanoparticles (NPs, address the potential food safety implications and health hazards connected with the consumption of nanofood. A number of health complications connected with the human exposure to ENMs are discussed, demonstrating that there is a real basis for the arisen concern not only connected with the gut health, but also with the potency to lead to systemic toxicity. The toxicological nature of hazard, exposure levels and risk to consumers from nanotechnology-derived food are on the earliest stage of investigation and this review also highlights the major gaps that need further research and regulation.

  12. Technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Selan, J.C.

    1981-07-01

    This report documents the technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1. The tests were to verify that faults on the non-Class 1E circuits would not propagate to the Class 1E circuits and degrade them below acceptable levels. The tests conducted demonstrated that the safety features actuation system did not degrade below acceptable levels nor was the system's ability to perform its protective functions affected

  13. From learning from accidents to teaching about accident causation and prevention: Multidisciplinary education and safety literacy for all engineering students

    International Nuclear Information System (INIS)

    Saleh, Joseph H.; Pendley, Cynthia C.

    2012-01-01

    In this work, we argue that system accident literacy and safety competence should be an essential part of the intellectual toolkit of all engineering students. We discuss why such competence should be taught and nurtured in engineering students, and provide one example for how this can be done. We first define the class of adverse events of interest as system accidents, distinct from occupational accidents, through their (1) temporal depth of causality and (2) diversity of agency or groups and individuals who influence or contribute to the accident occurrence/prevention. We then address the question of why the interest in this class of events and their prevention, and we expand on the importance of system safety literacy and the contributions that engineering students can make in the long-term towards accident prevention. Finally, we offer one model for an introductory course on accident causation and system safety, discuss the course logistics, material and delivery, and our experience teaching this subject. The course starts with the anatomy of accidents and is grounded in various case studies; these help illustrate the multidisciplinary nature of the subject, and provide the students with the important concepts to describe the phenomenology of accidents (e.g., initiating events, accident precursor or lead indicator, and accident pathogen). More importantly, the case studies invite a deep reflection on the underlying failure mechanisms, their generalizability, and the various safety levers for accident prevention. The course then proceeds to an exposition of defense-in-depth, safety barriers and principles, essential elements for an education in accident prevention, and it concludes with a presentation of basic concepts and tools for uncertainty and risk analysis. Educators will recognize the difficulties in designing a new course on such a broad subject. It is hoped that this work will invite comments and contributions from the readers, and that the journal will

  14. Adipose-Derived Stem Cells in Novel Approaches to Breast Reconstruction: Their Suitability for Tissue Engineering and Oncological Safety.

    Science.gov (United States)

    O'Halloran, Niamh; Courtney, Donald; Kerin, Michael J; Lowery, Aoife J

    2017-01-01

    Adipose-derived stem cells (ADSCs) are rapidly becoming the gold standard cell source for tissue engineering strategies and hold great potential for novel breast reconstruction strategies. However, their use in patients with breast cancer is controversial and their oncological safety, particularly in relation to local disease recurrence, has been questioned. In vitro, in vivo, and clinical studies using ADSCs report conflicting data on their suitability for adipose tissue regeneration in patients with cancer. This review aims to provide an overview of the potential role for ADSCs in breast reconstruction and to examine the evidence relating to the oncologic safety of their use in patients with breast cancer.

  15. Cyclic features of the consequences from a postulated nuclear accident: a case study of the third level probabilistic safety assessment

    International Nuclear Information System (INIS)

    Xinhe, LIU; Homma, Toshimitsu

    2002-01-01

    In the third level probabilistic safety assessment, one of the three popular meteorological sequence sampling methods is cyclic sampling. The rationale of cyclic sampling is obviously that cyclic variation is the significant characteristics of the meteorological sequences and the health consequences resulting from a postulated nuclear accident are also remarkably of cyclic features. In this work, a set of time series was established for different health consequences using S3 source term and a whole year meteorological data. OSCAAR software system was utilized in the calculation of the health consequences. It is shown by the analysis that diurnal variation is remarked for all the kinds of health consequences, implying that cyclic sampling would be more effective than random sampling. The results also showed that there are not any dominating frequencies in the spectra of the consequences so that cyclic sampling might be incompetent to reduce the third level PSA to a satisfied level. Therefore, new schemes of meteorological sampling should be developed in the light of consideration of complex coupling of meteorological condition and population distribution rather than the consideration of meteorological condition alone

  16. SAFETY

    CERN Multimedia

    Niels Dupont

    2013-01-01

    CERN Safety rules and Radiation Protection at CMS The CERN Safety rules are defined by the Occupational Health & Safety and Environmental Protection Unit (HSE Unit), CERN’s institutional authority and central Safety organ attached to the Director General. In particular the Radiation Protection group (DGS-RP1) ensures that personnel on the CERN sites and the public are protected from potentially harmful effects of ionising radiation linked to CERN activities. The RP Group fulfils its mandate in collaboration with the CERN departments owning or operating sources of ionising radiation and having the responsibility for Radiation Safety of these sources. The specific responsibilities concerning "Radiation Safety" and "Radiation Protection" are delegated as follows: Radiation Safety is the responsibility of every CERN Department owning radiation sources or using radiation sources put at its disposition. These Departments are in charge of implementing the requi...

  17. Economic benefits of safety-engineered sharp devices in Belgium - a budget impact model.

    Science.gov (United States)

    Hanmore, Emma; Maclaine, Grant; Garin, Fiona; Alonso, Alexander; Leroy, Nicolas; Ruff, Lewis

    2013-11-25

    Measures to protect healthcare workers where there is risk of injury or infection from medical sharps became mandatory in the European Union (EU) from May 2013. Our research objective was to estimate the net budget impact of introducing safety-engineered devices (SEDs) for prevention of needlestick injuries (NSIs) in a Belgian hospital. A 5-year incidence-based budget impact model was developed from the hospital inpatient perspective, comparing costs and outcomes with SEDs and prior-used conventional (non-safety) devices. The model accounts for device acquisition costs and costs of NSI management in 4 areas of application where SEDs are currently used: blood collection, infusion, injection and diabetes insulin administration. Model input data were sourced from the Institut National d'Assurance Maladie-Invalidité, published studies, clinical guidelines and market research. Costs are discounted at 3%. For a 420-bed hospital, 100% substitution of conventional devices by SEDs is estimated to decrease the cumulative 5-year incidence of NSIs from 310 to 75, and those associated with exposure to blood-borne viral diseases from 60 to 15. Cost savings from managing fewer NSIs more than offset increased device acquisition costs, yielding estimated 5-year overall savings of €51,710. The direction of these results is robust to a range of sensitivity and model scenario analyses. The model was most sensitive to variation in the acquisition costs of SEDs, rates of NSI associated with conventional devices, and the acquisition costs of conventional devices. NSIs are a significant potential risk with the use of sharp devices. The incidence of NSIs and the costs associated with their management can be reduced through the adoption of safer work practices, including investment in SEDs. For a Belgian hospital, the budget impact model reports that the incremental acquisition costs of SEDs are offset by the savings from fewer NSIs. The availability of more robust data for NSI reduction

  18. Quality assurance during construction of civil engineering structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2011-11-01

    The present manual is developed to deal with quality assurance aspect of civil engineering structures in a greater detail. This manual gives detail to develop QA plans specific to multifarious activities of civil engineering construction

  19. Engineered and Administrative Safety Systems for the Control of Prompt Radiation Hazards at Accelerator Facilities

    International Nuclear Information System (INIS)

    Liu, James C.; SLAC; Vylet, Vashek; Walker, Lawrence S.

    2007-01-01

    The ANSI N43.1 Standard, currently in revision (ANSI 2007), sets forth the requirements for accelerator facilities to provide adequate protection for the workers, the public and the environment from the hazards of ionizing radiation produced during and from accelerator operations. The Standard also recommends good practices that, when followed, provide a level of radiation protection consistent with those established for the accelerator communities. The N43.1 Standard is suitable for all accelerator facilities (using electron, positron, proton, or ion particle beams) capable of producing radiation, subject to federal or state regulations. The requirements (see word 'shall') and recommended practices (see word 'should') are prescribed in a graded approach that are commensurate with the complexity and hazard levels of the accelerator facility. Chapters 4, 5 and 6 of the N43.1 Standard address specially the Radiation Safety System (RSS), both engineered and administrative systems, to mitigate and control the prompt radiation hazards from accelerator operations. The RSS includes the Access Control System (ACS) and Radiation Control System (RCS). The main requirements and recommendations of the N43.1 Standard regarding the management, technical and operational aspects of the RSS are described and condensed in this report. Clearly some aspects of the RSS policies and practices at different facilities may differ in order to meet the practical needs for field implementation. A previous report (Liu et al. 2001a), which reviews and summarizes the RSS at five North American high-energy accelerator facilities, as well as the RSS references for the 5 labs (Drozdoff 2001; Gallegos 1996; Ipe and Liu 1992; Liu 1999; Liu 2001b; Rokni 1996; TJNAF 1994; Yotam et al. 1991), can be consulted for the actual RSS implementation at various laboratories. A comprehensive report describing the RSS at the Stanford Linear Accelerator Center (SLAC 2006) can also serve as a reference

  20. Early Engagement of Safety and Mission Assurance Expertise Using Systems Engineering Tools: A Risk-Based Approach to Early Identification of Safety and Assurance Requirements

    Science.gov (United States)

    Darpel, Scott; Beckman, Sean

    2016-01-01

    Decades of systems engineering practice have demonstrated that the earlier the identification of requirements occurs, the lower the chance that costly redesigns will needed later in the project life cycle. A better understanding of all requirements can also improve the likelihood of a design's success. Significant effort has been put into developing tools and practices that facilitate requirements determination, including those that are part of the model-based systems engineering (MBSE) paradigm. These efforts have yielded improvements in requirements definition, but have thus far focused on a design's performance needs. The identification of safety & mission assurance (S&MA) related requirements, in comparison, can occur after preliminary designs are already established, yielding forced redesigns. Engaging S&MA expertise at an earlier stage, facilitated by the use of MBSE tools, and focused on actual project risk, can yield the same type of design life cycle improvements that have been realized in technical and performance requirements.

  1. Recent Experiences of the NASA Engineering and Safety Center (NESC) GN and C Technical Discipline Team (TDT)

    Science.gov (United States)

    Dennehy, Cornelius J.

    2010-01-01

    The NASA Engineering and Safety Center (NESC), initially formed in 2003, is an independently funded NASA Program whose dedicated team of technical experts provides objective engineering and safety assessments of critical, high risk projects. The GN&C Technical Discipline Team (TDT) is one of fifteen such discipline-focused teams within the NESC organization. The TDT membership is composed of GN&C specialists from across NASA and its partner organizations in other government agencies, industry, national laboratories, and universities. This paper will briefly define the vision, mission, and purpose of the NESC organization. The role of the GN&C TDT will then be described in detail along with an overview of how this team operates and engages in its objective engineering and safety assessments of critical NASA projects. This paper will then describe selected recent experiences, over the period 2007 to present, of the GN&C TDT in which they directly performed or supported a wide variety of NESC assessments and consultations.

  2. Engineering and maintenance applied to safety-related valves in nuclear power plants

    International Nuclear Information System (INIS)

    Verdu, M. F.; Perez-Aranda, J.

    2014-01-01

    Nuclear Division in Iberdrola engineering and Construction has a team with extensive experience on engineering and services works related to valves. Also, this team is linked to UNESA as Technical support and Reference Center. Iberdrola engineering and construction experience in nuclear power plants valves, gives effective response to engineering and maintenance works that can be demanded in a nuclear power plant and it requires a high degree of qualification and knowledge both in Operation and Outages. (Author)

  3. Safety

    International Nuclear Information System (INIS)

    2001-01-01

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

  4. Automated System of Diagnostic Monitoring at Bureya HPP Hydraulic Engineering Installations: a New Level of Safety

    International Nuclear Information System (INIS)

    Musyurka, A. V.

    2016-01-01

    This article presents the design, hardware, and software solutions developed and placed in service for the automated system of diagnostic monitoring (ASDM) for hydraulic engineering installations at the Bureya HPP, and assuring a reliable process for monitoring hydraulic engineering installations. Project implementation represents a timely solution of problems addressed by the hydraulic engineering installation diagnostics section.

  5. Automated System of Diagnostic Monitoring at Bureya HPP Hydraulic Engineering Installations: a New Level of Safety

    Energy Technology Data Exchange (ETDEWEB)

    Musyurka, A. V., E-mail: musyurkaav@burges.rushydro.ru [Bureya HPP (a JSC RusGidro affiliate) (Russian Federation)

    2016-09-15

    This article presents the design, hardware, and software solutions developed and placed in service for the automated system of diagnostic monitoring (ASDM) for hydraulic engineering installations at the Bureya HPP, and assuring a reliable process for monitoring hydraulic engineering installations. Project implementation represents a timely solution of problems addressed by the hydraulic engineering installation diagnostics section.

  6. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-20

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D&D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities.

  7. Idaho National Engineering Laboratory (INEL) Environmental Restoration Program (ERP), Baseline Safety Analysis File (BSAF). Revision 1

    International Nuclear Information System (INIS)

    1994-01-01

    This document was prepared to take the place of a Safety Evaluation Report since the Baseline Safety Analysis File (BSAF)and associated Baseline Technical Safety Requirements (TSR) File do not meet the requirements of a complete safety analysis documentation. Its purpose is to present in summary form the background of how the BSAF and Baseline TSR originated and a description of the process by which it was produced and approved for use in the Environmental Restoration Program.The BSAF is a facility safety reference document for INEL environmental restoration activities including environmental remediation of inactive waste sites and decontamination and decommissioning (D ampersand D) of surplus facilities. The BSAF contains safety bases common to environmental restoration activities and guidelines for performing and documenting safety analysis. The common safety bases can be incorporated by reference into the safety analysis documentation prepared for individual environmental restoration activities with justification and any necessary revisions. The safety analysis guidelines in BSAF provide an accepted method for hazard analysis; analysis of normal, abnormal, and accident conditions; human factors analysis; and derivation of TSRS. The BSAF safety bases and guidelines are graded for environmental restoration activities

  8. Survey of materials and other problems of relevance in safety engineering, and an assessment of their reflection in regulatory guides for conventional and nuclear engineering (1. technical report)

    International Nuclear Information System (INIS)

    Trunk, M.; Herter, K.H.

    1984-01-01

    Survey and assessment of nuclear engineering specifications and regulatory guides (ASME-BPVC Section III, division 1,2 and KTA, BS 5500) with regard to materials, dimensioning and testing for the purpose of showing to what extent available technical codes, regulatory guides and safety codes are useful in preventing failures and defining the safe limit. The other question examined is that of how these codes ought to be brought up to date in order to reflect the latest state of the art in science and technology. (orig./HP) [de

  9. Safety evaluation report related to the operation of WPPSS Nuclear Project No. 2. Docket No. 50-397, Washington Public Power Supply System

    International Nuclear Information System (INIS)

    1982-08-01

    Information is presented concerning site characteristics; design criteria for structures, systems, and components; engineered safety features; instrumentation and control; auxiliary systems; conduct of operations; and financial qualifications

  10. Design, fabrication and erection of steel structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety Standard for Civil Engineering Structures Important to Safety of Nuclear Facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design, fabrication and erection of steel structures important to safety

  11. Pedestrian safety engineering and intelligent transportation system-based countermeasures program for reduced pedestrian fatalities, injuries, conflicts and other surrogate measures : Miami-Dade site.

    Science.gov (United States)

    2008-08-25

    This report presents the methods and key findings from the Miami-Dade comprehensive pedestrian safety planning and engineering project. It is one of three such projects in the nation funded by the Federal Highway Administration (FHWA) to evaluate: In...

  12. International cooperation in the safety and environmental assessment for the ITER engineering design activities

    International Nuclear Information System (INIS)

    Gordon, C.; Baker, D.J.; Bartels, H-W.

    1998-01-01

    The ITER Project includes design and assessment activities to ensure the safety and environmental attractiveness of ITER and demonstrate that it can be sited in any of the sponsoring Parties with a minimum of site-specific redesign. This paper highlights some of the efforts to develop an international consensus approach for ITER safety design and assessment, including: development of general safety and environmental design criteria; development of quantitative dose-release assessment criteria; development of a radiation protection program; waste characterization; and development of safety analysis guidelines. The high level of interaction, cooperation and collaboration between the Joint Central Team and the Home Teams, and between the safety team and designers, and the spirit of consensus that has guided them have resulted in a safe design for ITER and a safety design and assessment that can meet the needs of the potential host countries. (author)

  13. INNOVATIVE FORMS SUPPORTING SAFE METHODS OF WORK IN SAFETY ENGINEERING FOR THE DEVELOPMENT OF INTELLIGENT SPECIALIZATIONS

    Directory of Open Access Journals (Sweden)

    Anna GEMBALSKA-KWIECIEŃ

    2016-10-01

    Full Text Available The article discusses innovative forms of participation of employees in the work safety system. It also presents the advantages of these forms of employees’ involvement. The aim of empirical studies was the analysis of their behavior and attitude towards health and safety at work. The issues considered in the article have a significant impact on the improvement of methods of prevention related to work safety and aided the creation of a healthy society.

  14. ZIKA-001: Safety and Immunogenicity of an Engineered DNA Vaccine Against ZIKA virus infection

    Science.gov (United States)

    Tebas, Pablo; Roberts, Christine C; Muthumani, Kar; Reuschel, Emma; White, Scott; Khan, Amir S; Racine, Trina; Choi, Hyeree; Zaidi, Faraz; Boyer, Jean; Kudchodkar, Sagar; Park, Young K; Trottier, Sylvie; Remigio, Celine; Krieger, Diane; Kobinger, Gary P; Weiner, David; Maslow, Joel

    2017-01-01

    suggesting that the humoral response generated by the vaccine is protective in this model. Conclusion Our trial shows for the first time in humans the safety and immunogenicity of an engineered DNA encoding consensus viral protein against ZIKV. Future studies will evaluate the effectiveness of the vaccine. Disclosures C. C. Roberts, GeneOne: Member, Salary. S. White, GeneOne: Member, Salary. A. S. Khan, Inovio: Employee and Shareholder, Salary and Stock. J. Boyer, Inovio: Employee and Shareholder, Salary and Stock. Y. K. Park, GeneOne: Board Member, CEO and Employee, Salary and Stock. S. Trottier, Canadian Institutes of Health Research: Investigator, Research grant. C. Remigio, GeneOne: Employee and Shareholder, Salary and Stock. G. P. Kobinger, GeneOne: Grant Investigator and Scientific Advisor, Grant recipient and Research support. D. Weiner, GeneOne: Grant Investigator and Scientific Advisor, Grant recipient, Licensing agreement or royalty and Stock. J. Maslow, GeneOne: Employee, Salary and Stock.

  15. Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

    International Nuclear Information System (INIS)

    Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

    1996-01-01

    The rules that are currently under application to verify the acceptance of flaws in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation to reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R6 procedure for assessing the integrity of the structure. (authors). 5 refs., 5 figs

  16. Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

    International Nuclear Information System (INIS)

    Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

    1995-01-01

    The current rules applied to verify the flaws acceptance in nuclear components rely on deterministic criteria supposed to ensure the plant safe operation. The interest in have a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation do reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R 6 procedure for assessing the structure integrity. (author). 5 refs., 5 figs., 1 tab

  17. Procurement and quality control of components important to safety in nuclear engineering projects

    International Nuclear Information System (INIS)

    Zhang Zhihua; Zhang Yiyun

    2006-01-01

    The procurement and quality control of components is a very important work in the nuclear engineering. This paper introduces the project management techniques, such as how to make a plan of components purchase in nuclear engineering. This paper discussed the classification of components, evaluation of the potential suppliers, invitation of bids, exchange of design details with the suppliers, quality assurance and quality assurance audit, and the equipment checks before acceptance and some engineering experiences. (authors)

  18. Integration of Behaviour-Based Safety Programme into Engineering Laboratories and Workshops Conceptually

    Science.gov (United States)

    Koo, Kean Eng; Zain, Ahmad Nurulazam Md; Zainal, Siti Rohaida Mohamed

    2012-01-01

    The purpose of this conceptual research framework is to develop and integrate a safety training model using a behaviour-based safety training programme into laboratories for young adults, during their tertiary education, particularly in technical and vocational education. Hence, this research will be investigating the outcome of basic safety…

  19. Safety manual for civil engineering and building works of nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    This manual lays down the various important considerations that go into safe design of civil structures for nuclear power plants. This manual identifies the design approach, quality assurance requirements and acceptance criteria that need to be observed to assure safety. Considerations on civil design having bearing on safety, during decommissioning are also indicated. (original). 37 refs., tabs

  20. Systems Engineering Management Plan for Tank Farm Restoration and Safety Operations, Project W-314

    International Nuclear Information System (INIS)

    MCGREW, D.L.

    2000-01-01

    The Systems Engineering Management Plan for Project W-314 has been prepared within the guidelines of HNF-SD-WM-SEMP-002, TWRS Systems Engineering Management Plan. The activities within this SEMP have been tailored, in accordance with the TWRS SEMP and DOE Order 430.1, Life Cycle Asset Management, to meet the needs of the project

  1. Safety I-II, resilience and antifragility engineering: a debate explained through an accident occurring on a mobile elevating work platform.

    Science.gov (United States)

    Martinetti, Alberto; Chatzimichailidou, Maria Mikela; Maida, Luisa; van Dongen, Leo

    2018-04-24

    Occupational health and safety (OHS) represents an important field of exploration for the research community: in spite of the growth of technological innovations, the increasing complexity of systems involves critical issues in terms of degradation of the safety levels. In such a situation, new safety management approaches are now mandatory in order to face the safety implications of the current technological evolutions. Along these lines, performing risk-based analysis alone seems not to be enough anymore. The evaluation of robustness, antifragility and resilience of a socio-technical system is now indispensable in order to face unforeseen events. This article will briefly introduce the topics of Safety I and Safety II, resilience engineering and antifragility engineering, explaining correlations, overlapping aspects and synergies. Secondly, the article will discuss the applications of those paradigms to a real accident, highlighting how they can challenge, stimulate and inspire research for improving OHS conditions.

  2. Questioning the Role of Requirements Engineering in the Causes of Safety-Critical Software Failures

    Science.gov (United States)

    Johnson, C. W.; Holloway, C. M.

    2006-01-01

    Many software failures stem from inadequate requirements engineering. This view has been supported both by detailed accident investigations and by a number of empirical studies; however, such investigations can be misleading. It is often difficult to distinguish between failures in requirements engineering and problems elsewhere in the software development lifecycle. Further pitfalls arise from the assumption that inadequate requirements engineering is a cause of all software related accidents for which the system fails to meet its requirements. This paper identifies some of the problems that have arisen from an undue focus on the role of requirements engineering in the causes of major accidents. The intention is to provoke further debate within the emerging field of forensic software engineering.

  3. IAEA safety guides in the light of recent developments in earthquake engineering

    International Nuclear Information System (INIS)

    Gurpinar, A.

    1988-11-01

    The IAEA safety guides 50-SG-S1 and 50-SG-S2 emphasize on the determination of the design basis earthquake ground motion and earthquake resistant design considerations for nuclear power plants, respectively. Since the elaboration of these safety guides years have elapsed and a review of some of these concepts is necessary, taking into account the information collected and the technical developments. In this article, topics within the scope of these safety guides are discussed. In particular, the results of some recent research which may have a bearing on the nuclear industry are highlighted. Conclusions and recommendations are presented. 6 fig., 19 refs. (F.M.)

  4. Recommendations on the use of expert judgment in safety and reliability engineering studies. Two offshore case studies

    International Nuclear Information System (INIS)

    Hokstada, Per; Oien, Knut; Reinertsen, Rune

    1998-01-01

    This paper provides guidance on the process of establishing input data to safety and reliability engineering analyses when no or little field data exist, and expert judgment is required. Some recommendations are directly related to a discussion of basic requirements for scientific work. Further, two case studies are discussed in order to highlight some actual problem areas that are experienced when using expert judgment, and some recommendations for handling these problems are given. The first case describes how expert judgment was used to analyse the safe operation of an umbilical on a semisubmersible drilling rig, and the second case is related to establishing generic failure rates/probabilities for components of offshore safety systems

  5. Nuclear safety chains

    International Nuclear Information System (INIS)

    Robbins, M.C.; Eames, G.F.; Mayell, J.R.

    1981-01-01

    An original scheme has been developed for expressing the complex interrelationships associated with the engineered safeguards provided for a nuclear power station. This management tool, based upon network diagrams called Nuclear Safety Chains, looks at the function required of a particular item of safety plant, defines all of the vital supplies and support features necessary for successful operation, and expresses them in visual form, to facilitate analysis and optimisation for operations and maintenance staff. The safety chains are confined to manual schemes at present, although they are designed to be compatible with modern computer techniques. Their usefulness with any routine maintenance planning application on high technology plant is already being appreciated. (author)

  6. A new technology perspective and engineering tools approach for large, complex and distributed mission and safety critical systems components

    Science.gov (United States)

    Carrio, Miguel A., Jr.

    1988-01-01

    Rapidly emerging technology and methodologies have out-paced the systems development processes' ability to use them effectively, if at all. At the same time, the tools used to build systems are becoming obsolescent themselves as a consequence of the same technology lag that plagues systems development. The net result is that systems development activities have not been able to take advantage of available technology and have become equally dependent on aging and ineffective computer-aided engineering tools. New methods and tools approaches are essential if the demands of non-stop and Mission and Safety Critical (MASC) components are to be met.

  7. 77 FR 62224 - Hanford Tank Farms Flammable Gas Safety Strategy

    Science.gov (United States)

    2012-10-12

    ... (Board) believes that current operations at the Hanford Tank Farms require safety- significant active... administrative control in lieu of an engineered feature is also contrary to DOE's established hierarchy of...

  8. Food safety evaluation of crops produced through genetic engineering--how to reduce unintended effects?

    Science.gov (United States)

    Jelenić, Srećko

    2005-06-01

    Scientists started applying genetic engineering techniques to improve crops two decades ago; about 70 varieties obtained via genetic engineering have been approved to date. Although genetic engineering offers the most precise and controllable genetic modification of crops in entire history of plant improvement, the site of insertion of a desirable gene cannot be predicted during the application of this technology. As a consequence, unintended effects might occur due to activation or silencing of genes, giving rise to allergic reactions or toxicity. Therefore, extensive chemical, biochemical and nutritional analyses are performed on each new genetically engineered variety. Since the unintended effects may be predictable on the basis of what is known about the insertion place of the transgenic DNA, an important aim of plant biotechnology is to define techniques for the insertion of transgene into the predetermined chromosomal position (gene targeting). Although gene targeting cannot be applied routinely in crop plants, given the recent advances, that goal may be reached in the near future.

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

    Science.gov (United States)

    Zelkin, Natalie; Henriksen, Stephen

    2011-01-01

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

  10. The effectiveness of insurer-supported safety and health engineering controls in reducing workers' compensation claims and costs.

    Science.gov (United States)

    Wurzelbacher, Steven J; Bertke, Stephen J; Lampl, Michael P; Bushnell, P Timothy; Meyers, Alysha R; Robins, David C; Al-Tarawneh, Ibraheem S

    2014-12-01

    This study evaluated the effectiveness of a program in which a workers' compensation (WC) insurer provided matching funds to insured employers to implement safety/health engineering controls. Pre- and post-intervention WC metrics were compiled for the employees designated as affected by the interventions within 468 employers for interventions occurring from 2003 to 2009. Poisson, two-part, and linear regression models with repeated measures were used to evaluate differences in pre- and post-data, controlling for time trends independent of the interventions. For affected employees, total WC claim frequency rates (both medical-only and lost-time claims) decreased 66%, lost-time WC claim frequency rates decreased 78%, WC paid cost per employee decreased 81%, and WC geometric mean paid claim cost decreased 30% post-intervention. Reductions varied by employer size, specific industry, and intervention type. The insurer-supported safety/health engineering control program was effective in reducing WC claims and costs for affected employees. © 2014 Wiley Periodicals, Inc.

  11. History of nuclear power plants safety in France (1945-2000) - Engineer techniques, expert evaluation, topical issue

    International Nuclear Information System (INIS)

    Foasso, Cyrille

    2003-01-01

    This doctoral dissertation relates the history of the mastery of risks in civil nuclear plants in France. Since 1960, it's known as the 'surete nucleaire'. Over a fifty-year period separating the discovery or the atomic fission and its industrial application on a large scale this PhD shows which technical means were used over the years by engineers to handle this risk which is said to be huge. It also studies the various processes in expert evaluation and in decision making elaborated to evaluate if the risk was acceptable or not. Beyond the conflicts between nuclear advocates and opponents, this thesis shows how ever among nuclear engineer the growing distinction between roles (promoters, experts and controlling authorities) and the various jobs (designers, builders and plant operators) triggered different estimations as far as the methods to obtain a satisfactory safety. Thanks to the progress of knowledge through research programs, thanks to the lessons drawn from the functioning or dysfunction of nuclear plants, thanks to the reinforcement of regulations (which more or less reflects the public's opinion concerning this industry) the safety has progressively improved. Thus, this historical study is multiple: a technical history of technology, a history of scientific, industrial and administrative organization, a social history and finally an international and comparative history since the nuclear energy history quickly developed beyond national boundaries. (author) [fr

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  13. Study on safety performance evaluation system of nuclear engineering construction units based on AHP

    International Nuclear Information System (INIS)

    Xu Yulin; Sun Jian; Shi Xiaofan

    2012-01-01

    As a very effectual management mean, the performance management has extensively used by many companies of China for staff assessment. The author explored the establishment of the 'Safety Performance Evaluation System' by finding out the similarities in operation between a company and a team of nuclear power projects. Then the author analyzed the principles of the performance management and good practices and summarized safety management experiences. The weight of the system index by using AHP method was calculated in this article. (authors)

  14. Engineer Ethics

    International Nuclear Information System (INIS)

    Lee, Dae Sik; Kim, Yeong Pil; Kim, Yeong Jin

    2003-03-01

    This book tells of engineer ethics such as basic understanding of engineer ethics with history of engineering as a occupation, definition of engineering and specialized job and engineering, engineer ethics as professional ethics, general principles of ethics and its limitation, ethical theory and application, technique to solve the ethical problems, responsibility, safety and danger, information engineer ethics, biotechnological ethics like artificial insemination, life reproduction, gene therapy and environmental ethics.

  15. Engineered design features in the HI-STAR/HI-STORM systems to maximize ALARA, safety, and community acceptance

    International Nuclear Information System (INIS)

    Blessing, Christian

    2003-01-01

    Heltec International is a U.S. corporation headquartered in New Jersey, dedicated to providing capital goods and technical services to the power industry. Over 75 percent of the company's product output is destined for nuclear power plants. Holter counts among its active clients a majority of the nuclear plants in the United States, as well as Korea, Taiwan, Mexico, and Brazil. The company also has a growing market presence in Japan and the European Union. Leading U.S. nuclear plant owners, such as Entergy, Exelon, FPL, Southern Nuclear, PG and E and TVA have a long-term and continuous business relationship with Holtec International. This article deals with Holtec dry storage system description, the multi-purpose canister, hi-star 100 overpack, hi-storm 100 overpack and unique advantages of holtec's dry storage technology

  16. Risk and safety perception on urban and rural roads: Effects of environmental features, driver age and risk sensitivity.

    Science.gov (United States)

    Cox, Jolene A; Beanland, Vanessa; Filtness, Ashleigh J

    2017-10-03

    The ability to detect changing visual information is a vital component of safe driving. In addition to detecting changing visual information, drivers must also interpret its relevance to safety. Environmental changes considered to have high safety relevance will likely demand greater attention and more timely responses than those considered to have lower safety relevance. The aim of this study was to explore factors that are likely to influence perceptions of risk and safety regarding changing visual information in the driving environment. Factors explored were the environment in which the change occurs (i.e., urban vs. rural), the type of object that changes, and the driver's age, experience, and risk sensitivity. Sixty-three licensed drivers aged 18-70 years completed a hazard rating task, which required them to rate the perceived hazardousness of changing specific elements within urban and rural driving environments. Three attributes of potential hazards were systematically manipulated: the environment (urban, rural); the type of object changed (road sign, car, motorcycle, pedestrian, traffic light, animal, tree); and its inherent safety risk (low risk, high risk). Inherent safety risk was manipulated by either varying the object's placement, on/near or away from the road, or altering an infrastructure element that would require a change to driver behavior. Participants also completed two driving-related risk perception tasks, rating their relative crash risk and perceived risk of aberrant driving behaviors. Driver age was not significantly associated with hazard ratings, but individual differences in perceived risk of aberrant driving behaviors predicted hazard ratings, suggesting that general driving-related risk sensitivity plays a strong role in safety perception. In both urban and rural scenes, there were significant associations between hazard ratings and inherent safety risk, with low-risk changes perceived as consistently less hazardous than high

  17. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    Science.gov (United States)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  18. Using human factors engineering to improve patient safety in the cardiovascular operating room.

    Science.gov (United States)

    Gurses, Ayse P; Martinez, Elizabeth A; Bauer, Laura; Kim, George; Lubomski, Lisa H; Marsteller, Jill A; Pennathur, Priyadarshini R; Goeschel, Chris; Pronovost, Peter J; Thompson, David

    2012-01-01

    Despite significant medical advances, cardiac surgery remains a high risk procedure. Sub-optimal work system design characteristics can contribute to the risks associated with cardiac surgery. However, hazards due to work system characteristics have not been identified in the cardiovascular operating room (CVOR) in sufficient detail to guide improvement efforts. The purpose of this study was to identify and categorize hazards (anything that has the potential to cause a preventable adverse patient safety event) in the CVOR. An interdisciplinary research team used prospective hazard identification methods including direct observations, contextual inquiry, and photographing to collect data in 5 hospitals for a total 22 cardiac surgeries. We performed thematic analysis of the qualitative data guided by a work system model. 60 categories of hazards such as practice variations, high workload, non-compliance with evidence-based guidelines, not including clinicians' in medical device purchasing decisions were found. Results indicated that hazards are common in cardiac surgery and should be eliminated or mitigated to improve patient safety. To improve patient safety in the CVOR, efforts should focus on creating a culture of safety, increasing compliance with evidence based infection control practices, improving communication and teamwork, and designing better tools and technologies through partnership among all stakeholders.

  19. Environment, Safety and Health progress assessment of the Idaho National Engineering Laboratory (INEL)

    International Nuclear Information System (INIS)

    1993-08-01

    The ES ampersand H Progress Assessments are part of the Department's continuous improvement process throughout DOE and its contractor organizations. The purpose of the INEL ES ampersand H Progress Assessment is to provide the Department with concise independent information on the following: (1) change in culture and attitude related to ES ampersand H activities; (2) progress and effectiveness of the ES ampersand H corrective actions resulting from previous Tiger Team Assessments; (3) adequacy and effectiveness of the ES ampersand H self-assessment programs of the DOE line organizations and the site management and operating contractor; and (4) effectiveness of DOE and contractor management structures, resources, and systems to effectively address ES ampersand H problems. It is not intended that this Progress Assessment be a comprehensive compliance assessments of ES ampersand H activities. The points of reference for assessing programs at the INEL were, for the most part, the 1991 INEL Tiger Team Assessment, the INEL Corrective Action Plan, and recent appraisals and self-assessments of INEL. Horizontal and vertical reviews of the following programmatic areas were conducted: Management: Corrective action program; self-assessment; oversight; directives, policies, and procedures; human resources management; and planning, budgeting, and resource allocation. Environment: Air quality management, surface water management, groundwater protection, and environmental radiation. Safety and Health: Construction safety, worker safety and OSHA, maintenance, packaging and transportation, site/facility safety review, and industrial hygiene

  20. Human factors science and safety engineering : can the STAMP model serve in establishing a common language?

    NARCIS (Netherlands)

    Karanikas, Nektarios; Schwarz, M; Harfmann, J

    2017-01-01

    A symbiotic relationship between human factors and safety scientists is needed to ensure the provision of holistic solutions for problems emerging in modern socio-technical systems. System Theoretic Accident Model and Processes (STAMP) tackles both interactions and individual failures of human and

  1. Risk evaluation method for faults by engineering approach. (1) Nuclear safety for accident scenario and measures for fault movement

    International Nuclear Information System (INIS)

    Narabayashi, Tadashi; Chiba, Go; Okamoto, Koji; Kameda, Hiroyuki; Ebisawa, Katsumi; Yamazaki, Haruo; Konagai, Kazuo; Kamiya, Masanobu; Nagasawa, Kazuyuki

    2016-01-01

    Japan, as a frequent earthquake country, has a responsibility to resolve efficient measures to enhance nuclear safety, to continue utilizing the nuclear power, based on the risks and importance levels in the scientific and rational manner. In his paper describes how to evaluate the risk of faults movement by engineering approach. An open fruitful discussion by experts in the various area of earthquake, geology, geotechnical, civil, and a seismic design as well as other stakeholders such as academia professors, nuclear reactor engineers, regulators, and licensees. The Atomic Energy Society established an Investigation Committee on Development of Activity and Risk Evaluation Method for Faults by Engineering Approach (IC-DAREFEA) on October 1st, a 2014. The Investigation Committee utilizes the most advanced scientific and rational judgement, and continuous discussions and efforts in the global field, in order to collect and organize these knowledge and reflect the global standards and nuclear regulations, such as risk evaluation method for the faults movements and prevention of severe accidents, based on the accumulated database in the world, including Chuetsuoki Earthquake, North Nagano Earthquake and Kumamoto Earthquake. (author)

  2. Building Models to Better Understand the Importance of Cost versus Safety in Engineering

    Science.gov (United States)

    Sumrall, William; Mott, Michael

    2010-01-01

    While some disasters involving engineered structures are due to events in nature (e.g., tornadoes, hurricanes, earthquakes), others may be caused by inadequate materials, design flaws, and poor maintenance. These catastrophes result in the loss of human lives and cost billions of dollars. In the set of lessons described here, students design a…

  3. The engineer's anxiety about plant safety as viewed by the lawyer

    International Nuclear Information System (INIS)

    Lendi, M.

    1981-01-01

    The role of the design and operating engineers is described. The responsibilities of management is stressed and the legal and moral aspects of the problem are emphasized. The law can only give guidance but the ultimate responsibility rests with the design and operating staff. (G.R.S.)

  4. A comprehensive framework for evaluating the environmental health and safety implications of engineered nanomaterials

    Science.gov (United States)

    Engineered nanomaterials (ENM) are a growing aspect of the global economy, and their safe and sustainable development, use and eventual disposal requires the capability to forecast and avoid potential problems. This review is concerned with the releases of ENM into the environmen...

  5. SAFETY

    CERN Multimedia

    M. Plagge, C. Schaefer and N. Dupont

    2013-01-01

    Fire Safety – Essential for a particle detector The CMS detector is a marvel of high technology, one of the most precise particle measurement devices we have built until now. Of course it has to be protected from external and internal incidents like the ones that can occur from fires. Due to the fire load, the permanent availability of oxygen and the presence of various ignition sources mostly based on electricity this has to be addressed. Starting from the beam pipe towards the magnet coil, the detector is protected by flooding it with pure gaseous nitrogen during operation. The outer shell of CMS, namely the yoke and the muon chambers are then covered by an emergency inertion system also based on nitrogen. To ensure maximum fire safety, all materials used comply with the CERN regulations IS 23 and IS 41 with only a few exceptions. Every piece of the 30-tonne polyethylene shielding is high-density material, borated, boxed within steel and coated with intumescent (a paint that creates a thick co...

  6. SAFETY

    CERN Multimedia

    C. Schaefer and N. Dupont

    2013-01-01

      “Safety is the highest priority”: this statement from CERN is endorsed by the CMS management. An interpretation of this statement may bring you to the conclusion that you should stop working in order to avoid risks. If the safety is the priority, work is not! This would be a misunderstanding and misinterpretation. One should understand that “working safely” or “operating safely” is the priority at CERN. CERN personnel are exposed to different hazards on many levels on a daily basis. However, risk analyses and assessments are done in order to limit the number and the gravity of accidents. For example, this process takes place each time you cross the road. The hazard is the moving vehicle, the stake is you and the risk might be the risk of collision between both. The same principle has to be applied during our daily work. In particular, keeping in mind the general principles of prevention defined in the late 1980s. These principles wer...

  7. Safety profile and long-term engraftment of human CD31+ blood progenitors in bone tissue engineering.

    Science.gov (United States)

    Zigdon-Giladi, Hadar; Elimelech, Rina; Michaeli-Geller, Gal; Rudich, Utai; Machtei, Eli E

    2017-07-01

    Endothelial progenitor cells (EPCs) participate in angiogenesis and induce favorable micro-environments for tissue regeneration. The efficacy of EPCs in regenerative medicine is extensively studied; however, their safety profile remains unknown. Therefore, our aims were to evaluate the safety profile of human peripheral blood-derived EPCs (hEPCs) and to assess the long-term efficacy of hEPCs in bone tissue engineering. hEPCs were isolated from peripheral blood, cultured and characterized. β tricalcium phosphate scaffold (βTCP, control) or 10 6 hEPCs loaded onto βTCP were transplanted in a nude rat calvaria model. New bone formation and blood vessel density were analyzed using histomorphometry and micro-computed tomography (CT). Safety of hEPCs using karyotype analysis, tumorigenecity and biodistribution to target organs was evaluated. On the cellular level, hEPCs retained their karyotype during cell expansion (seven passages). Five months following local hEPC transplantation, on the tissue and organ level, no inflammatory reaction or dysplastic change was evident at the transplanted site or in distant organs. Direct engraftment was evident as CD31 human antigens were detected lining vessel walls in the transplanted site. In distant organs human antigens were absent, negating biodistribution. Bone area fraction and bone height were doubled by hEPC transplantation without affecting mineral density and bone architecture. Additionally, local transplantation of hEPCs increased blood vessel density by nine-fold. Local transplantation of hEPCs showed a positive safety profile. Furthermore, enhanced angiogenesis and osteogenesis without mineral density change was found. These results bring us one step closer to first-in-human trials using hEPCs for bone regeneration. Copyright © 2017 International Society for Cellular Therapy. Published by Elsevier Inc. All rights reserved.

  8. Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.N.; Mohan, L.R.

    2002-01-01

    Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

  9. Human Factors engineering criteria and design for the Hanford Waste Vitrification Plant preliminary safety analysis report

    International Nuclear Information System (INIS)

    Wise, J.A.; Schur, A.; Stitzel, J.C.L.

    1993-09-01

    This report provides a rationale and systematic methodology for bringing Human Factors into the safety design and operations of the Hanford Waste Vitrification Plant (HWVP). Human Factors focuses on how people perform work with tools and machine systems in designed settings. When the design of machine systems and settings take into account the capabilities and limitations of the individuals who use them, human performance can be enhanced while protecting against susceptibility to human error. The inclusion of Human Factors in the safety design of the HWVP is an essential ingredient to safe operation of the facility. The HWVP is a new construction, nonreactor nuclear facility designed to process radioactive wastes held in underground storage tanks into glass logs for permanent disposal. Its design and mission offer new opposites for implementing Human Factors while requiring some means for ensuring that the Human Factors assessments are sound, comprehensive, and appropriately directed

  10. Communication practices in engineering, manufacturing, and research for food and water safety

    CERN Document Server

    2015-01-01

    This book demonstrates some of the ways in which communication and developing technologies can improve global food and water safety by providing a historical background on outbreaks and public resistance, as well as generating interest in youth and potential professionals in the field : History of muckraking in the food industry ; Case study on groundwater regulation ; Interviews with members of the beef industry and livestock market owners.

  11. Army Corps of Engineers: Actions Needed to Improve Cost Sharing for Dam Safety Repairs

    Science.gov (United States)

    2015-12-01

    agreements with the Corps, their history of being a sponsor, the financial impacts of cost sharing for dam safety repair projects, and the Corps...1240 (2007)) and Beaver Lake dam, AR (Pub. L. No. 102-377, 106 Stat. 1315, 1318 (1992), Pub. L. No. 102-580, § 209(f), 106 Stat. 4797, 4830 (1992...inaction in setting a clear policy for a provision under which sponsors face significant financial impacts has contributed to conditions under

  12. ORSERG resource book. Operational reactor safety engineering and review group. Final report, March 1992

    International Nuclear Information System (INIS)

    1992-03-01

    EPRI has prepared this resource book to help utilities with their Self-Assessment Programs at nuclear power plants. Self-assessments are reviews performed by nuclear power plant utilities to monitor plant performance status and adequacy, identify trends in operational activities important to safety, and assess the impact of these trends on plant safety. Activities performed as self-assessments include reviews and evaluations of plant performance and abnormal events, technical evaluations of plant activities to identify potential problem areas, and reviews of other sources of plant design and operating experience for applicability to safety. This resource book is based on information obtained from utilities and includes examples of activities and methods that have proven effective. The resource book includes a summary of NRC requirements, guidelines for self-assessment program planning, descriptions and examples of investigative techniques, and key references that can be consulted for additional information. It can serve as a training guide for plant staff members who are assigned to self-assessment activities. (author)

  13. Application in nuclear engineering: methodology of innovative nuclear reactors: approaches to the safety of future nuclear power plants

    International Nuclear Information System (INIS)

    Alramady, A.M.K

    2008-01-01

    This thesis describes RELAP5 and MATLAB/SIMULINK computer codes for thermal hydraulic analysis of a typical pressurized water reactor (PWR). The two codes are used to calculate the thermal-hydraulic characteristics of the reactor core and the primary loop under steady-state and hypothetical accidents conditions.New designs of nuclear power plants are directed to increase safety by many methods like reducing the dependence on active parts (such as safety pumps, fans, and diesel generators ) and replacing them with passive features such as gravity draining of cooling water from tanks, and natural circulation of water and air. In this work, high and medium pressure injection pumps are replaced by passive injection components. Different break sizes in cold leg pipe are simulated to analyze to what degree the plant is safe (without any operator action) by using only these passive components. The passive design means operators would not need to take immediate action after an accident, with the reactor ,instead, safely shutting down on its own. Different accident scenarios were simulated in this thesis as loss of coolant accidents and station blackout accidents, and complete passive safety systems used to mitigate theses accidents.

  14. Engineered Nanomaterials in Food: Implications for Food Safety and Consumer Health

    OpenAIRE

    Martirosyan, Alina; Schneider, Yves-Jacques

    2014-01-01

    From the current state-of-the-art, it is clear that nanotechnology applications are expected to bring a range of benefits to the food sector aiming at providing better quality and conservation. In the meantime, a growing number of studies indicate that the exposure to certain engineered nanomaterials (ENMs) has a potential to lead to health complications and that there is a need for further investigations in order to unravel the biological outcomes of nanofood consumption. In the current revi...

  15. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    Rankin, D.B.

    1984-01-01

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  16. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical

  17. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical.

  18. Mixed and low-level waste treatment project: Appendix C, Health and safety criteria for the mixed and low-level waste treatment facility at the Idaho National Engineering Laboratory. Part 2, Chemical constituents

    Energy Technology Data Exchange (ETDEWEB)

    Neupauer, R.M.; Thurmond, S.M.

    1992-09-01

    This report contains health and safety information relating to the chemicals that have been identified in the mixed waste streams at the Waste Treatment Facility at the Idaho National Engineering Laboratory. Information is summarized in two summary sections--one for health considerations and one for safety considerations. Detailed health and safety information is presented in material safety data sheets (MSDSs) for each chemical.

  19. Safety

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    Aspects of fission reactors are considered - control, heat removal and containment. Brief descriptions of the reactor accidents at the SL-1 reactor (1961), Windscale (1957), Browns Ferry (1975), Three Mile Island (1979) and Chernobyl (1986) are given. The idea of inherently safe reactor designs is discussed. Safety assessment is considered under the headings of preliminary hazard analysis, failure mode analysis, event trees, fault trees, common mode failure and probabalistic risk assessments. These latter can result in a series of risk distributions linked to specific groups of fault sequences and specific consequences. A frequency-consequence diagram is shown. Fatal accident incidence rates in different countries including the United Kingdom for various industries are quoted. The incidence of fatal cancers from occupational exposure to chemicals is tabulated. Human factors and the acceptability of risk are considered. (U.K.)

  20. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M; Boncheva, H; Stafanov, D [Central Laboratory for Seismic Mechanics and Earthquake Engineering, Bulgarian Academy of Sciences, Sofia (Bulgaria)

    1993-07-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation.