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Sample records for line valve accident

  1. Technical evaluation: 300 Area steam line valve accident

    International Nuclear Information System (INIS)

    1993-08-01

    On June 7, 1993, a journeyman power operator (JPO) was severely burned and later died as a result of the failure of a 6-in. valve that occurred when he attempted to open main steam supply (MSS) valve MSS-25 in the U-3 valve pit. The pit is located northwest of Building 331 in the 300 Area of the Hanford Site. Figure 1-1 shows a layout of the 300 Area steam piping system including the U-3 steam valve pit. Figure 1-2 shows a cutaway view of the approximately 10- by 13- by 16-ft-high valve pit with its various steam valves and connecting piping. Valve MSS-25, an 8-in. valve, is located at the bottom of the pit. The failed 6-in. valve was located at the top of the pit where it branched from the upper portion of the 8-in. line at the 8- by 8- by 6-in. tee and was then ''blanked off'' with a blind flange. The purpose of this technical evaluation was to determine the cause of the accident that led to the failure of the 6-in. valve. The probable cause for the 6-in. valve failure was determined by visual, nondestructive, and destructive examination of the failed valve and by metallurgical analysis of the fractured region of the valve. The cause of the accident was ultimately identified by correlating the observed failure mode to the most probable physical phenomenon. Thermal-hydraulic analyses, component stress analyses, and tests were performed to verify that the probable physical phenomenon could be reasonably expected to produce the failure in the valve that was observed

  2. Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

    International Nuclear Information System (INIS)

    Yang, C. Y.; Bang, Y. S.; Kim, H. J.

    2001-01-01

    A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

  3. Butterfly valve of all rubber lining type

    International Nuclear Information System (INIS)

    Shimada, Shosaku; Nakatsuma, Sumiya; Sasaki, Iwao; Aoki, Naoshi.

    1982-01-01

    The valves used for the circulating water pipes for condensers in nuclear and thermal power stations have become large with the increase of power output, and their specifications have become strict. The materials for the valves change from cast iron to steel plate construction. To cope with sea water corrosion, rubber lining has been applied to the internal surfaces of valve boxes, and the build-up welding of stainless steel has been made on the edges of valves. However, recently it is desired to develop butterfly valves, of which the whole valve disks are lined with hard rubber. For the purpose of confirming the performance of large bore valves, a 2600 mm bore butterfly valve of all rubber lining type was used, and the opening and closing test of 1100 times was carried out by applying thermal cycle and pressure difference and using artifical sea water. Also the bending test of hard rubber lining was performed with test pieces. Thus, it was confirmed that the butterfly valves of all rubber lining type have the performance exceeding that of the valves with build-up welding. The course of development of the valves of all rubber lining type, the construction and the items of confirmation by tests of these valves, and the tests of the valve and the hard rubber lining described above are reported. (Kako, I.)

  4. Transcatheter aortic valve implantation and cerebrovascular accidents.

    Science.gov (United States)

    Stortecky, Stefan; Wenaweser, Peter; Windecker, Stephan

    2012-09-01

    Transcatheter aortic valve implantation (TAVI) is an evidence-based treatment alternative for selected high-risk patients with symptomatic severe aortic stenosis as acknowledged in the most recent edition of the ESC Guidelines on Valvular Heart Disease 2012. However, periprocedural complications and in particular cerebrovascular accidents remain a matter of concern. While transcatheter heart valve technology continuously improves and the development of novel and even less invasive implantation techniques is on-going, cerebrovascular events complicating TAVI may abrogate the usual improvement in terms of prognosis and quality of life. This article describes the incidence of cerebrovascular events after cardiovascular procedures, provides an overview of the pathophysiological mechanisms as well as the impact on outcomes and provides some insights into preventive strategies as well as the acute management of these events.

  5. An on-line pressurizer surveillance system design to prevent small-break loss-of-coolant accidents through power-operated relief valves using a microcomputer

    International Nuclear Information System (INIS)

    Lee, J.H.; Chang, S.H.

    1987-01-01

    A small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve is one of the important contributors to nuclear power plant risk. A pressurizer surveillance system was designed to use a microcomputer to prevent the malfunction of the system; the effect of this improvement has been assessed through probabilistic risk assessment. The microcomputer diagnoses the malfunction of the system by a process-checking method and automatically performs the backup action related to each malfunction. This improvement means that we can correctly diagnose ''spurious opening,'' ''failure to reclose,'' and ''small-break LOCA,'' which are difficult for operators to diagnose quickly and correctly, and by taking automatic backup action one can reduce the probability of human error

  6. Valve for closing a steam line

    International Nuclear Information System (INIS)

    Meyer, W.; Potrykus, G.

    1976-01-01

    Instead of several control elements, the quick-closing valve, especially in the main-steam line between steam generator and turbine of a power station has the valve cone itself as the only movable part, acting with its inner surface as a piston within a second cylinder space. The valve shaft is at the same time a piston rod with a stepped piston at the upper end. This piston is loaded in a cylinder at the upspace below the valve cover on one hand by a spring, on the other hand by its own medium. Two non-return valves, one of it in a bore of the valve cone, connect the first-mentioned cylinder space with the steam-loaded inlet resp. outlet side of the valve. For controlling the valve, a magnet valve is sufficient. By automatic control of the valve cone coupled with several pistons several control lines can be omitted. There are also no pressurized control lines outside the valve which could be damaged by exterior influences. (ERA) [de

  7. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  8. Coupled simulation of steam line break accident

    International Nuclear Information System (INIS)

    Royer, E.; Raimond, E.; Caruge, D.

    2000-01-01

    The steam line break is a PWR type reactor design accident, which concerns coupled physical phenomena. To control these problems simulation are needed to define and validate the operating procedures. The benchmark OECD PWR MSLB (Main Steam Line Break) has been proposed by the OECD to validate the feasibility and the contribution of the multi-dimensional tools in the simulation of the core transients. First the benchmark OECD PWR MSLB is presented. Then the analysis of the three exercises (system with pinpoint kinetic, three-dimensional core and whole system with three-dimensional core) are discussed. (A.L.B.)

  9. Study of typical nuclear containment purge valves in an accident environment

    International Nuclear Information System (INIS)

    Watkins, J.C.; Steele, R. Jr.; Hill, R.C.; DeWall, K.G.

    1986-08-01

    This report presents the results of the containment purge and vent valve test program, conducted under the sponsorship of the United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research. The test program investigated butterfly valve operability and leak integrity under light-water-reactor design basis and severe accident conditions. Three nuclear-designed butterfly valves typical of those used in domestic nuclear power plant containment purge and vent applications were tested. For a comparison of response, two valve of the same size with differing internal designs were tested. For extrapolation insights, a larger-sized valve similar to one of the smaller valves was also tested. Dynamic flow tests were performed over the range of design basis accident pressures. Leak integrity testing was also performed at both design basis and severe accident temperatures and pressures. The valve experiments were performed with various piping configurations and valve orientations to the flow to simulate the various installation options in field applications. Testing was also performed in a standard ANSI test section

  10. Integral isolation valve systems for loss of coolant accident protection

    Science.gov (United States)

    Kanuch, David J.; DiFilipo, Paul P.

    2018-03-20

    A nuclear reactor includes a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel having vessel penetrations that exclusively carry flow into the nuclear reactor and at least one vessel penetration that carries flow out of the nuclear reactor. An integral isolation valve (IIV) system includes passive IIVs each comprising a check valve built into a forged flange and not including an actuator, and one or more active IIVs each comprising an active valve built into a forged flange and including an actuator. Each vessel penetration exclusively carrying flow into the nuclear reactor is protected by a passive IIV whose forged flange is directly connected to the vessel penetration. Each vessel penetration carrying flow out of the nuclear reactor is protected by an active IIV whose forged flange is directly connected to the vessel penetration. Each active valve may be a normally closed valve.

  11. Incidence of cerebrovascular accidents in patients undergoing minimally invasive valve surgery.

    Science.gov (United States)

    LaPietra, Angelo; Santana, Orlando; Mihos, Christos G; DeBeer, Steven; Rosen, Gerald P; Lamas, Gervasio A; Lamelas, Joseph

    2014-07-01

    Minimally invasive valve surgery has been associated with increased cerebrovascular complications. Our objective was to evaluate the incidence of cerebrovascular accidents in patients undergoing minimally invasive valve surgery. We retrospectively reviewed all the minimally invasive valve surgery performed at our institution from January 2009 to June 2012. The operative times, lengths of stay, postoperative complications, and mortality were analyzed. A total of 1501 consecutive patients were identified. The mean age was 73 ± 13 years, and 808 patients (54%) were male. Of the 1501 patients, 206 (13.7%) had a history of a cerebrovascular accident, and 225 (15%) had undergone previous heart surgery. The procedures performed were 617 isolated aortic valve replacements (41.1%), 658 isolated mitral valve operations (43.8%), 6 tricuspid valve repairs (0.4%), 216 double valve surgery (14.4%), and 4 triple valve surgery (0.3%). Femoral cannulation was used in 1359 patients (90.5%) and central cannulation in 142 (9.5%). In 1392 patients (92.7%), the aorta was clamped, and in 109 (7.3%), the surgery was performed with the heart fibrillating. The median aortic crossclamp and cardiopulmonary bypass times were 86 minutes (interquartile range [IQR], 70-107) minutes and 116 minutes (IQR, 96-143), respectively. The median intensive care unit length of stay was 47 hours (IQR, 29-74), and the median postoperative hospital length of stay was 7 days (IQR, 5-10). A total of 23 cerebrovascular accidents (1.53%) and 38 deaths (2.53%) had occurred at 30 days postoperatively. Minimally invasive valve surgery was associated with an acceptable stroke rate, regardless of the cannulation technique. Copyright © 2014 The American Association for Thoracic Surgery. Published by Mosby, Inc. All rights reserved.

  12. Neutronic calculations for Angra-1 steam line break accident

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Sato, Sadakatu

    2000-01-01

    The reduction of boron concentration in the Boron Injection Tank (BIT), to the room temperature solubility level, makes necessary a reanalysis of the steam line break accident of Angra 1 NPP. This paper describes the neutronic calculation related to this reanalysis. The main steps of the work were: review of reactivity parameters used in the accident simulation; search of xenon profiles that cause the most severe core power distribution; calculation of hot channel factors and other neutronic parameters necessary for DNBR determination. The final conclusion, related to the steam line break accident, states the BIT concentration may be reduced to 2000 ppm. (author)

  13. The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

    2005-01-01

    The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

  14. Analysis of the loss of coolant accident due to the faiture in the open position of two pressurizer relief valves, for Angra-1 nuclear power plant

    International Nuclear Information System (INIS)

    Freire, C.F.

    1981-06-01

    A study of the modeling techniques adequate for simulating the loss of coolant accident caused by stuck open pressurizer relief valves, using the RELAP4-MOD5 code, is performed and the model developed is applied to the analysis of this kind of accident for the Central Nuclear Almirante Alvaro Alberto Unit (Angra 1). The thermal hydraulic behavior of the reactor cooling system, when subjected to a loss of main feedwater followed by the failure in the open position of two pressurizer relief valves, is determined. The relief valves are assumed to fail in the totally open position, delivering the maximum massflow through the discharge line. The RELAP4-MOD5 code is shown to be adequate for this kind of analysis, and the detailed prediction of the thermal hydraulic behavior of the Reactor Coolant System is thus possible. The eficiency of the emergency core cooling system of Angra 1 is demonstrated, the fuel elements remaining covered by the coolant during all the accident, and the peak clad temperatures are kept within design limites, ensuring the integrity of the core. (Author) [pt

  15. Cerebrovascular accidents complicating transcatheter aortic valve implantation: frequency, timing and impact on outcomes.

    Science.gov (United States)

    Stortecky, Stefan; Windecker, Stephan; Pilgrim, Thomas; Heg, Dik; Buellesfeld, Lutz; Khattab, Ahmed A; Huber, Christoph; Gloekler, Steffen; Nietlispach, Fabian; Mattle, Heinrich; Jüni, Peter; Wenaweser, Peter

    2012-05-15

    Cerebrovascular accidents (CVA) are considered among the most serious adverse events after transcatheter aortic valve implantation (TAVI). The objective of the present study was to evaluate the frequency and timing of CVA after TAVI and to investigate the impact on clinical outcomes within 30 days of the procedure. Between August 2007 and October 2011, 389 high-risk elderly patients with symptomatic severe aortic stenosis underwent TAVI via transfemoral, transapical or subclavian access. A total of 14 patients (3.6%) experienced at least one CVA within 30 days of follow-up and most events (74%) occurred within the first day of the procedure. Patients with CVA had an increased risk of all-cause (42.3% vs. 5.1%, ORadjusted 11.7, 95% CI 3.4-40.3, pCerebrovascular accidents among patients undergoing TAVI occur predominantly during the periprocedural period, are associated with multiple implantation attempts of the bioprosthesis and significantly impair prognosis.

  16. Analysis of containment parameters during the main steam line break with the failure of the feedwater control valves

    International Nuclear Information System (INIS)

    Fabjan, L.; Petelin, S.; Mavko, B.; Gortnar, O.; Tiselj, I.

    1992-01-01

    U.S. Nuclear Regulatory Commission (NRC) information notice 91-69: 'Errors in Main Steam Line Break Analyses for Determining Containment Parameters' shows the possibility of an accident which could lead to beyond design containment pressure and temperature. Such accident would be caused by the continuation of feedwater flow following a main stream line break (MSLB) inside the containment. Krsko power plant already experienced problems with main feedwater control valves. For that reason, analysis of MSLB has been performed taking into account continuous feedwater addition scenario and different containment safety systems capabilities availability. Steam and water released into the containment during MSLB was calculated using RELAP5/MOD2 computer code. The containment response to MSLB was calculated using CONTEMPT-LT/028 computer code. The results indicated that the continuous feedwater flow following a MSLB could lead to beyond design containment pressure. The peak pressure and temperature depend on isolation time for main- and auxiliary-feedwater supply. In the case of low boron concentration injection, the core recriticality is characteristic for this type of accidents. It was concluded that the presented analysis of MSLB with continuous feedwater addition scenario is the worst case for containment design

  17. Recent experience with testing of parallel disc gate valves under accident flow conditions

    International Nuclear Information System (INIS)

    LaPointe, P.A.; Clayton, J.K.

    1992-01-01

    This paper presents the nuclear valve industry's latest and most extensive valve qualification test program experience. The test program includes a variety of 25 different gate and globe valves. All the test valves are power operated using either air, electric, or gas/hydraulic operators. The valves are categorized in size and pressure class so as to form a group of appropriate parent valve assemblies. Parent valve assembly qualification is used as the basis for qualification of candidate valve assemblies. The parent and candidate valve assemblies are representative of a nuclear plant's safety-related valve applications. The test program was performed in accordance with ANSI B16.41-1983 'Functional Qualification Requirements for Power Operated Active Valve Assemblies for Nuclear Power Plants.' The focus of this paper is on functional valve qualification test experience and specifically flow interruption testing to Annex G of the aforementioned test standard. Results of the flow test are summarized, including the coefficient of friction for each of the gate type valves reported. Information on valve size, pressure class, and actuator are given for all valves in the program. Although all valves performed extremely well, only selected test data are presented. The effects of the speed of operation and the effects of different fluid flow rates as they relate to the coefficient of friction between the valve disc and seat are discussed. The variation in the coefficient of friction based on other variables in the thrust equation, namely, differential pressure area is cited

  18. On-line valve monitoring at the Ormen Lange gas plant

    Energy Technology Data Exchange (ETDEWEB)

    Greenlees, R.; Hale, S. [Score Atlanta Inc., Kennesaw, Georgia (United States)

    2011-07-01

    The purpose of this presentation is to discuss replacing time and labor intensive nuclear outage activities with on line condition monitoring solutions, primarily the periodic verification of MOV functionality discussed in USNRC Generic Letter 96.05. This regulation requires that MOV age related performance degradations are properly identified and accounted for, causing utilities to have to retest valves periodically for the duration of the plants operating license. AECL designed CANDU reactors have a world class performance and safety record, with typical average annual capacity factors of 90%. The CANDU reactor design has the ability to refuel on line, as a result (a) it can be a challenge scheduling all required valve testing into limited duration outage work windows, (b) at multi unit sites, Unit 0 valves can be difficult to test because they are rarely ever out of service, (c) deuterium-oxide (heavy water) moderator is expensive to manufacture, as a result, effective through valve leakage monitoring is essential. These three factors alone make CANDU sites the most suitable candidates for on line valve monitoring systems. Nuclear industry regulations have been instrumental in the development of 'at the valve' diagnostic systems, but diagnostic testing has not typically been utilized to the same degree in other less regulated industries. However, that trend is changing, and the move toward valve diagnostics and condition monitoring has moved fastest in the offshore oil and gas industry on the Norwegian side of the North Sea. The Ormen Lange plant, located on Nyhamna Island on the west coast of Norway, operated by Shell, is one of the worlds most advanced gas processing plants. A stated maintenance goal for the plant is that 70% of the maintenance budget and spend should be based on the results of on line condition monitoring, utilizing monitoring systems equipped with switch sensing, strain gages, hydraulic and pneumatic pressure transducers and

  19. On-line valve monitoring at the Ormen Lange gas plant

    International Nuclear Information System (INIS)

    Greenlees, R.; Hale, S.

    2011-01-01

    The purpose of this presentation is to discuss replacing time and labor intensive nuclear outage activities with on line condition monitoring solutions, primarily the periodic verification of MOV functionality discussed in USNRC Generic Letter 96.05. This regulation requires that MOV age related performance degradations are properly identified and accounted for, causing utilities to have to retest valves periodically for the duration of the plants operating license. AECL designed CANDU reactors have a world class performance and safety record, with typical average annual capacity factors of 90%. The CANDU reactor design has the ability to refuel on line, as a result (a) it can be a challenge scheduling all required valve testing into limited duration outage work windows, (b) at multi unit sites, Unit 0 valves can be difficult to test because they are rarely ever out of service, (c) deuterium-oxide (heavy water) moderator is expensive to manufacture, as a result, effective through valve leakage monitoring is essential. These three factors alone make CANDU sites the most suitable candidates for on line valve monitoring systems. Nuclear industry regulations have been instrumental in the development of 'at the valve' diagnostic systems, but diagnostic testing has not typically been utilized to the same degree in other less regulated industries. However, that trend is changing, and the move toward valve diagnostics and condition monitoring has moved fastest in the offshore oil and gas industry on the Norwegian side of the North Sea. The Ormen Lange plant, located on Nyhamna Island on the west coast of Norway, operated by Shell, is one of the worlds most advanced gas processing plants. A stated maintenance goal for the plant is that 70% of the maintenance budget and spend should be based on the results of on line condition monitoring, utilizing monitoring systems equipped with switch sensing, strain gages, hydraulic and pneumatic pressure transducers and acoustic leakage

  20. On-line diagnostic techniques for air-operated control valves based on time series analysis

    International Nuclear Information System (INIS)

    Ito, Kenji; Matsuoka, Yoshinori; Minamikawa, Shigeru; Komatsu, Yasuki; Satoh, Takeshi.

    1996-01-01

    The objective of this research is to study the feasibility of applying on-line diagnostic techniques based on time series analysis to air-operated control valves - numerous valves of the type which are used in PWR plants. Generally the techniques can detect anomalies by failures in the initial stages for which detection is difficult by conventional surveillance of process parameters measured directly. However, the effectiveness of these techniques depends on the system being diagnosed. The difficulties in applying diagnostic techniques to air-operated control valves seem to come from the reduced sensitivity of their response as compared with hydraulic control systems, as well as the need to identify anomalies in low level signals that fluctuate only slightly but continuously. In this research, simulation tests were performed by setting various kinds of failure modes for a test valve with the same specifications as of a valve actually used in the plants. Actual control signals recorded from an operating plant were then used as input signals for simulation. The results of the tests confirmed the feasibility of applying on-line diagnostic techniques based on time series analysis to air-operated control valves. (author)

  1. Selection and evaluation of an ultra high vacuum gate valve for Isabelle beam line vacuum system

    International Nuclear Information System (INIS)

    Foerster, C.L.; McCafferty, D.

    1980-01-01

    A minimum of eighty-four (84) Ultra High Vacuum Gate Valves will be utilized in ISABELLE to protect proton beam lines from catastrophic vacuum failure and to provide sector isolation for maintenance requirements. The valve to be selected must function at less than 1 x 10 -11 Torr pressure and be bakeable to 300 0 C in its open or closed position. In the open position, the valve must have an RF shield to make the beam line walls appear continuous. Several proposed designs were built and evaluated. The evaluation consisted mainly of leak testing, life tests, thermal cycling, mass spectrometer analysis, and 10 -12 Torr operation. Problems with initial design and fabrication were resolved. Special requirements for design and construction were developed. This paper describes the tests on two final prototypes which appear to be the best candidates for ISABELLE operation

  2. Analysis of acetal toilet fill valve supply line nut failure

    Directory of Open Access Journals (Sweden)

    Anthony Timpanaro

    2017-10-01

    Full Text Available In recent years, there has been a rise in the number of product liability cases involving the failure of toilet water supply line acetal plastic nuts. These nuts can fail in service, causing water leaks that result in significant property and financial losses. This study examines three possible failure modes of acetal plastic toilet water supply nuts. The three failure modes tested were all due to over load failure of the acetal nut and are as follows: (1 Overtightening of the supply line acetal nut, (2 Supply line lateral pull and, (3 Embrittled supply line lateral pull. Additionally, a “hand-tight” torque survey was conducted. The fracture surfaces and characteristics of these failure tests were examined with Stereo Microscopy and Scanning Electron Microscopy (SEM. The failure modes were compared and contrasted to provide guidance in determination of cause in these investigations.

  3. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

  4. Severe Accident Management System On-line Network SAMSON

    International Nuclear Information System (INIS)

    Silverman, Eugene B.

    2004-01-01

    SAMSON is a computational tool used by accident managers in the Technical Support Centers (TSC) and Emergency Operations Facilities (EOF) in the event of a nuclear power plant accident. SAMSON examines over 150 status points monitored by nuclear power plant process computers during a severe accident and makes predictions about when core damage, support plate failure, and reactor vessel failure will occur. These predictions are based on the current state of the plant assuming that all safety equipment not already operating will fail. SAMSON uses expert systems, as well as neural networks trained with the back propagation learning algorithms to make predictions. Training on data from an accident analysis code (MAAP - Modular Accident Analysis Program) allows SAMSON to associate different states in the plant with different times to critical failures. The accidents currently recognized by SAMSON include steam generator tube ruptures (SGTRs), with breaks ranging from one tube to eight tubes, and loss of coolant accidents (LOCAs), with breaks ranging from 0.0014 square feet (1.30 cm 2 ) in size to breaks 3.0 square feet in size (2800 cm 2 ). (author)

  5. Short-course thrombolysis as the first line of therapy for cardiac valve thrombosis.

    Science.gov (United States)

    Manteiga, R; Carlos Souto, J; Altès, A; Mateo, J; Arís, A; Dominguez, J M; Borrás, X; Carreras, F; Fontcuberta, J

    1998-04-01

    To retrospectively evaluate the clinical and echocardiographic criteria of thrombolytic therapy for mechanical heart valve thrombosis. Nineteen consecutive patients with 22 instances of prosthetic heart valve thrombosis (14 mitral, 2 aortic, 3 tricuspid, and 3 pulmonary) were treated with short-course thrombolytic therapy as first option of treatment in absence of contraindications. The thrombolytic therapy protocol consisted of streptokinase (1,500,000 IU in 90 minutes) (n = 18) in one (n = 7) or two (n = 11) cycles or recombinant tissue-type plasminogen activator (100 mg in 90 minutes) (n = 4). Overall success was seen in 82%, immediate complete success in 59%, and partial success in 23%. Six patients without total response to thrombolytic therapy underwent surgery, and pannus was observed in 83%. Six patients showed complications: allergy, stroke, transient ischemic attack, coronary embolism, minor bleeding, and one death. At diagnosis, 10 patients evidenced atrial thrombus by transesophageal echocardiography, 3 of whom experienced peripheral embolism during thrombolysis. Four episodes of rethrombosis were observed (16%). The survivorship was 84% with a mean follow-up of 42.6 months. A short-course of thrombolytic therapy may be considered first-line therapy for prosthetic heart valve thrombosis. The risk of peripheral embolism may be evaluated for the presence of atrial thrombus by transesophageal echocardiography at diagnosis.

  6. Helium pressures in RHIC vacuum cryostats and relief valve requirements from magnet cooling line failure

    Energy Technology Data Exchange (ETDEWEB)

    Liaw, C.J.; Than, Y.; Tuozzolo, J.

    2011-03-28

    A catastrophic failure of the RHIC magnet cooling lines, similar to the LHC superconducting bus failure incident, would pressurize the insulating vacuum in the magnet and transfer line cryostats. Insufficient relief valves on the cryostats could cause a structural failure. A SINDA/FLUINT{reg_sign} model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the vacuum cryostat and discharging via the reliefs into the RHIC tunnel, had been developed to calculate the helium pressure inside the cryostat. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces were included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Existing relief valve sizes were reviewed to make sure that the maximum stresses, caused by the calculated maximum pressures inside the cryostats, did not exceed the allowable stresses, based on the ASME Code B31.3 and ANSYS results. The conclusions are as follows: (1) The S/F simulation results show that the highest internal pressure in the cryostats, due to the magnet line failure, is {approx}37 psig (255115 Pa); (2) Based on the simulation, the temperature on the cryostat chamber, INJ Q8-Q9, could drop to 228 K, which is lower than the material minimum design temperature allowed by the Code; (3) Based on the ASME Code and ANSYS results, the reliefs on all the cryostats inside the RHIC tunnel are adequate to protect the vacuum chambers when the magnet cooling lines fail; and (4) In addition to the pressure loading, the thermal deformations, due to the temperature decrease on the cryostat chambers, could also cause a high stress on the chamber, if not properly supported.

  7. Helium pressures in RHIC vacuum cryostats and relief valve requirements from magnet cooling line failure

    International Nuclear Information System (INIS)

    Liaw, C.J.; Than, Y.; Tuozzolo, J.

    2011-01-01

    A catastrophic failure of the RHIC magnet cooling lines, similar to the LHC superconducting bus failure incident, would pressurize the insulating vacuum in the magnet and transfer line cryostats. Insufficient relief valves on the cryostats could cause a structural failure. A SINDA/FLUINT(reg s ign) model, which simulated the 4.5K/4 atm helium flowing through the magnet cooling system distribution lines, then through a line break into the vacuum cryostat and discharging via the reliefs into the RHIC tunnel, had been developed to calculate the helium pressure inside the cryostat. Arc flash energy deposition and heat load from the ambient temperature cryostat surfaces were included in the simulations. Three typical areas: the sextant arc, the Triplet/DX/D0 magnets, and the injection area, had been analyzed. Existing relief valve sizes were reviewed to make sure that the maximum stresses, caused by the calculated maximum pressures inside the cryostats, did not exceed the allowable stresses, based on the ASME Code B31.3 and ANSYS results. The conclusions are as follows: (1) The S/F simulation results show that the highest internal pressure in the cryostats, due to the magnet line failure, is ∼37 psig (255115 Pa); (2) Based on the simulation, the temperature on the cryostat chamber, INJ Q8-Q9, could drop to 228 K, which is lower than the material minimum design temperature allowed by the Code; (3) Based on the ASME Code and ANSYS results, the reliefs on all the cryostats inside the RHIC tunnel are adequate to protect the vacuum chambers when the magnet cooling lines fail; and (4) In addition to the pressure loading, the thermal deformations, due to the temperature decrease on the cryostat chambers, could also cause a high stress on the chamber, if not properly supported.

  8. Effects of non-latching blast valves on the source term and consequences of the design-basis accidents in the Device Assembly Facility (DAF)

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1993-08-01

    The analysis of the Design-Basis Accidents (DBA) involving high explosives (HE) and Plutonium (Pu) in the assembly cell of the Device Assembly Facility (DAF), which was completed earlier, assumed latching blast valves in the ventilation system of the assembly cell. Latching valves effectively sealed a release path through the ventilation duct system. However, the blast valves in the assembly cell, as constructed are actually non-latching valves, and would reopen when the gas pressure drops to 0.5 psi above one atmosphere. Because the reopening of the blast valves provides an additional release path to the environment, and affects the material transport from the assembly cell to other DAF buildings, the DOE/NV DAF management has decided to support an additional analysis of the DAF's DBA to account for the effects of non-latching valves. Three cases were considered in the DAF's DBA, depending on the amount of HE and Pu involved, as follows: Case 1 -- 423 number-sign HE, 16 kg Pu; Case 2 -- 150 number-sign HE 10 kg Pu; Case 3 -- 55 number-sign HE 5 kg Pu. The results of the analysis with non-latching valves are summarized

  9. The tightness of the globe valves in the exploitations practice of the gas pipe-lines

    International Nuclear Information System (INIS)

    Pietrak, T.; Rudzki, Z.; Surmacz, W.

    2006-01-01

    Technological units of the Transit Gas Pipeline (i.e. Compressor Stations, Valve Stations, Stations or National Network Service Installations) have been fitted with Ball Valves as shut-off devices (block valves). Internal tightness of the valves' seat becomes major factor in securing proper service conditions during normal pipeline operation as well as for isolating of pipeline sections in emergency situations (loss of pipeline integrity or uncontrolled gas escape). Internal tightness of the valves is being inspected during scheduled maintenance of the pipeline units. Any leak revealed during inspection is being repaired, following instructions provided in the Manufacturer's Valve Manual. After a time, some cases have been identified, when repair of the revealed leak was found to be difficult, despite close following of the repair manuals. The paper presents analysis of the issue and corrective actions taken accordingly. (authors)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. On-line measurement of gaseous iodine species during a PWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Haykal, I.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, 91191 Gif sur Yvette Cedex, (France); Perrin, A. [CNRS-University of Paris Est and Paris 7, Laboratoire Inter-Universitaire des Systemes Atmospheriques, 94010 Creteil, (France); Vincent, B. [University of Burgundy, Laboratoire de physique, CNRS UMR 5027, 9, Avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Manceron, L. [Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); Mejean, G. [University of Joseph Fourier in Grenoble, Laboratoire de Spectrometrie Physique-CNRS UMR 5588, 38402 Saint Martin d' Heres, (France); Ducros, G. [CEA Cadarache, CEA, DEN, Departement d' Etudes des Combustibles, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    A long-range remote sensing of severe accidents in nuclear power plants can be obtained by monitoring the online emission of volatile fission products such as xenon, krypton, caesium and iodine. The nuclear accident in Fukushima was ranked at level 7 of the International Nuclear Event Scale by the NISA (Nuclear and Industrial Safety Agency) according to the importance of the radionuclide release and the off-site impact. Among volatile fission products, iodine species are of high concern, since they can be released under aerosols as well as gaseous forms. Four years after the Fukushima accident, the aerosol/gaseous partition is still not clear. Since the iodine gaseous forms are less efficiently trapped by the Filtered Containment Venting Systems than aerosol forms, it is of crucial importance to monitor them on-line during a nuclear accident, in order to improve the source term assessment in such a situation. Therefore, we propose to detect and quantify these iodine gaseous forms by the use of highly sensitive optical methods. (authors)

  12. Analysis of Two Electrocution Accidents in Greece that Occurred due to Unexpected Re-energization of Power Lines

    Directory of Open Access Journals (Sweden)

    Aikaterini D. Baka

    2014-09-01

    Full Text Available Investigation and analysis of accidents are critical elements of safety management. The over-riding purpose of an organization in carrying out an accident investigation is to prevent similar accidents, as well as seek a general improvement in the management of health and safety. Hundreds of workers have suffered injuries while installing, maintaining, or servicing machinery and equipment due to sudden re-energization of power lines. This study presents and analyzes two electrical accidents (1 fatal injury and 1 serious injury that occurred because the power supply was reconnected inadvertently or by mistake.

  13. Analysis of Two Electrocution Accidents in Greece that Occurred due to Unexpected Re-energization of Power Lines.

    Science.gov (United States)

    Baka, Aikaterini D; Uzunoglu, Nikolaos K

    2014-09-01

    Investigation and analysis of accidents are critical elements of safety management. The over-riding purpose of an organization in carrying out an accident investigation is to prevent similar accidents, as well as seek a general improvement in the management of health and safety. Hundreds of workers have suffered injuries while installing, maintaining, or servicing machinery and equipment due to sudden re-energization of power lines. This study presents and analyzes two electrical accidents (1 fatal injury and 1 serious injury) that occurred because the power supply was reconnected inadvertently or by mistake.

  14. IE Information Notice No. 85-47: Potential effect of line-induced vibration on certain Target Rock solenoid-operated valves

    International Nuclear Information System (INIS)

    Jordan, E.L.

    1992-01-01

    On November 14, 1984, Arizona Public Services Company provided the NRC with a final report on a 10 CFR 50.55(e) reportable condition relating to qualification testing of certain TR (Target Rock), solenoid-operated valves. Four TR valves, procured by Combustion Engineering (CE) for use at Palo Verde Nuclear Generating Station Unit 3, were tested to the requirements of NUREG-0588, Category 1. Test valves included two 1-inch TR valves, model 77L-001 and two 2-inch TR valves, model 77L-003. The qualification test involved irradiation to 50 megarads, thermal aging at 260 F for 635 hours, mechanical cycling, vibrational aging to represent normal service vibration, seismic testing, and finally, testing in a simulated LOCA environment. The licensee reported that during the qualification testing, a number of anomalies were identified, and the test was discontinued when the test valves failed to function for different reasons during the seismic testing. CE an TR appraised the overall safety significance of the observed test anomalies for the licensee. They considered the failure of the valve to open on demand as a result of solenoid lead shorting caused by line-induced vibrational wear to be a common mode of failure that, in a seismic event, could potentially disable several redundant valves at the same time. This failure of the valve to open on demand is the only observed test anomaly considered to have significant generic safety implications and is the subject of this information notice

  15. Source term analysis for a criticality accident in metal production line glove boxes

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1991-06-01

    A recent development in criticality accident analysis is the deterministic calculations of the transport of fission products and actinides through the barriers of the physical facility. The knowledge of the redistribution of the materials inside the facility will help determine the reentry and clean-up procedures. The amount of radioactive materials released to the environment is the source term for dispersion calculations. We have used an integrated computer model to determine the release of fission products to the environment from a hypothetical criticality event in a glove box of the metal production line (MPL) at the Lawrence Livermore National Laboratory (LLNL)

  16. Pressure disequilibria induced by rapid valve closure in noble gas extraction lines

    Science.gov (United States)

    Morgan, Leah; Davidheiser-Kroll, Brett

    2015-01-01

    Pressure disequilibria during rapid valve closures can affect calculated molar quantities for a range of gas abundance measurements (e.g., K-Ar geochronology, (U-Th)/He geochronology, noble gas cosmogenic chronology). Modeling indicates this effect in a system with a 10 L reservoir reaches a bias of 1% before 1000 pipette aliquants have been removed from the system, and a bias of 10% before 10,000 aliquants. Herein we explore the causes and effects of this problem, which is the result of volume changes during valve closure. We also present a solution in the form of an electropneumatic pressure regulator that can precisely control valve motion. This solution reduces the effect to ∼0.3% even after 10,000 aliquants have been removed from a 10 L reservoir.

  17. Pressure disequilibria induced by rapid valve closure in noble gas extraction lines

    Science.gov (United States)

    Morgan, Leah E.; Davidheiser-Kroll, Brett

    2015-06-01

    Pressure disequilibria during rapid valve closures can affect calculated molar quantities for a range of gas abundance measurements (e.g., K-Ar geochronology, (U-Th)/He geochronology, noble gas cosmogenic chronology). Modeling indicates this effect in a system with a 10 L reservoir reaches a bias of 1% before 1000 pipette aliquants have been removed from the system, and a bias of 10% before 10,000 aliquants. Herein we explore the causes and effects of this problem, which is the result of volume changes during valve closure. We also present a solution in the form of an electropneumatic pressure regulator that can precisely control valve motion. This solution reduces the effect to ˜0.3% even after 10,000 aliquants have been removed from a 10 L reservoir.

  18. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    International Nuclear Information System (INIS)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

    2001-01-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  19. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  20. On-line measurements of RuO{sub 4} during a PWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Reymond-Laruinaz, S.; Doizi, D. [CEA, DEN, Departement de Physico-chimie, CEA/Saclay, 91191 Gif sur Yvette Cedex, (France); Manceron, L. [Societe Civile Synchrotron SOLEIL, L' Orme des Merisiers, St-Aubin BP48, 91192 Gif-sur-Yvette Cedex, (France); MONARIS, UMR 8233, Universite Pierre et Marie Curie, 4 Place Jussieu, case 49, F-75252 Paris Cedex 05, (France); Boudon, V. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-Universite de Bourgogne, 9 avenue Alain Savary, BP 47870, F-21078 Dijon Cedex, (France); Ducros, G. [CEA, DEN, Departement d' Etudes des Combustibles, CEA/Cadarache, 13108 Saint-Paul-lez-Durance cedex, (France)

    2015-07-01

    After the Fukushima accident, it became essential to have a way to monitor in real time the evolution of a nuclear reactor during a severe accident, in order to react efficiently and minimize the industrial, ecological and health consequences of the accident. Among gaseous fission products, the tetroxide of ruthenium RuO{sub 4} is of prime importance since it has a significant radiological impact. Ruthenium is a low volatile fission product but in case of the rupture of the vessel lower head by the molten corium, the air entering into the vessel oxidizes Ru into gaseous RuO{sub 4}, which is not trapped by the Filtered Containment Venting Systems. To monitor the presence of RuO{sub 4} allows making a diagnosis of the core degradation and quantifying the release into the atmosphere. To determine the presence of RuO{sub 4}, FTIR spectrometry was selected. To study the feasibility of the monitoring, high-resolution IR measurements were realized at the French synchrotron facility SOLEIL on the infrared beam line AILES. Thereafter, theoretical calculations were done to simulate the FTIR spectrum to describe the specific IR fingerprint of the molecule for each isotope and based on its partial pressure in the air. (authors)

  1. Recommendations for main line block valves installation in gas pipelines; Recomendacoes para instalacao de valvulas de bloqueio de linha tronco em gasodutos

    Energy Technology Data Exchange (ETDEWEB)

    Frisoli, Caetano [TRANSPETRO - PETROBRAS Transportes, Rio de Janeiro, RJ (Brazil); Oliveira, Valeriano Duque de [PETROBRAS, Rio de Janeiro, RJ (Brazil)

    2003-07-01

    Cases of gas pipelines block valves and its pneumatic actuators presenting problems during the final pipeline commissioning and pre-operation phases, like internal leaks, leaking to the atmosphere, pneumatic circuit defects caused by water and debris, are nearly common. The majority can be avoided if a series of measuring are to be planned and implemented, as well as if an adequate planning of commissioning operations and line gasification, valves and actuators, are to be applied. This paper shows the practical experience in the construction and commissioning of valves and its actuators in the Bolivia-Brazil gas pipeline, which, in the first construction phase had a series of problems. After the diagnosis a set of procedures was implemented in the secondary construction phase, resulting in insignificant problems detected. All measures and procedures taken in the planning process, as well as additional aspects related to the main line valve design, its by-passes and supports, are demonstrated. (author)

  2. Analytical model for computing transient pressures and forces in the safety/relief valve discharge line. Mark I Containment Program, task number 7.1.2

    International Nuclear Information System (INIS)

    Wheeler, A.J.

    1978-02-01

    An analytical model is described that computes the transient pressures, velocities and forces in the safety/relief valve discharge line immediately after safety/relief valve opening. Equations of motion are defined for the gas-flow and water-flow models. Results are not only verified by comparing them with an earlier version of the model, but also with Quad Cities and Monticello plant data. The model shows reasonable agreement with the earlier model and the plant data

  3. SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

    2012-10-15

    APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

  4. Generation and Characterization of Vascular Smooth Muscle Cell Lines Derived from a Patient with a Bicuspid Aortic Valve

    Directory of Open Access Journals (Sweden)

    Pamela Lazar-Karsten

    2016-04-01

    Full Text Available Thoracic aortic dilation is the most common malformation of the proximal aorta and is responsible for 1%–2% of all deaths in industrialized countries. In approximately 50% of patients with a bicuspid aortic valve (BAV, dilation of any or all segments of the aorta occurs. BAV patients with aortic dilation show an increased incidence of cultured vascular smooth muscle cell (VSMC loss. In this study, VSMC, isolated from the ascending aorta of BAV, was treated with Simian virus 40 to generate a BAV-originated VSMC cell line. To exclude any genomic DNA or cross-contamination, highly polymorphic short tandem repeats of the cells were profiled. The cells were then characterized using flow cytometry and karyotyping. The WG-59 cell line created is the first reported VSMC cell line isolated from a BAV patient. Using an RT2 Profiler PCR Array, genes within the TGFβ/BMP family that are dependent on losartan treatment were identified. Endoglin was found to be among the regulated genes and was downregulated in WG-59 cells following treatment with different losartan concentrations, when compared to untreated WG-59 cells.

  5. Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1998-01-01

    Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

  6. Core to surge-line energy transport in a severe accident scenario

    International Nuclear Information System (INIS)

    Marzo, M. di; Almenas, K.; Gopalnarayanan, S.

    1994-01-01

    The analysis of loss of coolant accidents in a nuclear power plant, which progress to the stage where the core is uncovered, poses important safety related questions. One of these concerns the rate of energy transport to metal components of the primary system. An experimental program has been conducted at the Univ. of Maryland test facility which quantifies the rate of energy transfer from an uncovered core in a B ampersand W (once-through type steam generators) plant. SF 6 is used to simulate the natural circulation driving force of the high pressure steam expected at prototypical conditions. A time-dependent scaling methodology is developed to transpose experimental data to prototypical conditions. To achieve this transformation, a nominal fluid temperature increase rate of 1.0 degrees C/s is inferred from available TMI-2 event data. To bracket the range of potential prototypical transient scenarios, temperature ramps of 0.8 degrees C/s and 1.2 degrees C/s are also considered. Repeated tests, covering a range of test facility conditions, lead to estimated failure times at the surge line nozzle of 1.5 to 2 hours after initiation of the natural circulation phase of the transient

  7. Nuclear valves latest development

    International Nuclear Information System (INIS)

    Isaac, F.; Monier, M.

    1993-01-01

    In the frame of Nuclear Power Plant upgrade (Emergency Power Supply and Emergency Core Cooling), Westinghouse had to face a new valve design philosophy specially for motor operated valves. The valves have to been designed to resist any operating conditions, postulated accident or loss of control. The requirements for motor operated valves are listed and the selected model and related upgrading explained. As part of plant upgrade and valves replacement, Westinghouse has sponsored alternative hardfacing research programme. Two types of materials have been investigated: nickel base alloys and iron base alloys. Programme requirements and test results are given. A new globe valve model (On-Off or regulating) is described developed by Alsthom Velan permitting the seat replacement in less than 10 min. (Z.S.) 2 figs

  8. Redesign of emergency water supply system by-pass line from Cernavoda NPP Unit 1 and 2 using self regulating valves

    International Nuclear Information System (INIS)

    Tenescu, Mircea; Bigu, Melania; Nita, Iulian Pavel

    2010-01-01

    In this paper one considered improving the EWS (emergency water supply system) by-pass line in order to replace current manual operated valve with an self operated valve. This change is necessary in order to reduce the human error events in operation of the system in case of a DBE (design basis earthquake). The paper describes a theoretical and practical operation of a system using self regulating flow rate valves. Basically, the elimination of a possible human error in operating a system is important for nuclear safety in case of a DBE because it makes it avoidable in normal reactor cooling systems. The paper describes checking of the functioning of this equipment in operating conditions, investigating how it responds to various operating regimes. (authors)

  9. On the use of dental ceramics as a possible second-line approach to accident irradiation dosimetry

    International Nuclear Information System (INIS)

    Davies, J.E.

    1979-01-01

    Recent development in dental ceramic production has resulted in natural or depleted uranium, used for over half a century to mimic the fluorescence of natural teeth, being substituted in such ceramics by non-radioactive fluorescent materials. This creates the possibility of using dental ceramics incorporating the latter as second-line dosimeters in cases of accidental irradiation. This pilot study shows the feasbility of such an approach using both thermally stimulated exoelectron and thermoluminescent techniques. In conclusion, it is considered that it would be of interest to continue this investigation of dental ceramic materials as second-line accident dosimeters

  10. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    International Nuclear Information System (INIS)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2012-01-01

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described

  11. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    Energy Technology Data Exchange (ETDEWEB)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

  12. On-line fission products measurements during a PWR severe accident: the French DECA-PF project

    Energy Technology Data Exchange (ETDEWEB)

    Ducros, G.; Allinei, P.G.; Roure, C. [CEA, DEN, F-13108 Saint-Paul-lez-Durance, (France); Rozel, C. [EDF SEPTEN, 12-14 Avenue Dutrievoz, F-69628, Villeurbanne, (France); Blanc De Lanaute, N. [CANBERRA, 1 rue des Herons, F-78182, Saint Quentin en Yvelines, (France); Musoyan, G. [AREVA, Tour AREVA, 1 place Jean Millier, F-92084 Paris La Defense Cedex, (France)

    2015-07-01

    Following the Fukushima accident, a lot of recommendations was drawn by international organizations (IAEA, OECD, NUGENIA network...) in order to improve the safety in such accidental conditions and mitigate their consequences. One of these recommendations was to improve the robustness of the instrumentation, which was dramatically lacking at Fukushima, as well as to better determine the Source Term involved in nuclear accident. The DECA-PF project (Diagnosis of a degraded reactor core through Fission Product measurements) was elaborated in this context and selected as one of 21 collaborative R and D projects in the field of nuclear safety and radioprotection, funded in May 2013 by the French National Research Agency. Over the months following the Fukushima accident, a CEA crisis team was held in order to analyze on-line the situation taking into account the data delivered by TEPCO and other organizations. Despite the difficulties encountered concerning the reliability of these data, the work performed showed the high capacity of Fission Products (FP) measurements to get a diagnosis relative to the status of the reactors and the spent fuel pools (SFP). Based on these FP measurements, it was possible to conclude that the main origin of the releases was coming from the cores and not from the SFP, in particular for SFP-4 which was of high concern, and that the degradation level of the reactors was very large, including probably an extensive core melting. To improve the reliability of this kind of diagnosis, the necessity to get such measurements as soon as possible after the accident and as near as possible from the reactor was stressed. In this way the present DECA-PF project intends to develop a new and innovative instrumentation taking into account the design of the French nuclear power plants on which sand bed filters have been implemented for severe accident management. Three complementary techniques, devoted to measure the FP release on-line, are being studied

  13. Heart valve surgery

    Science.gov (United States)

    ... replacement; Valve repair; Heart valve prosthesis; Mechanical valves; Prosthetic valves ... surgery. Your heart valve has been damaged by infection ( endocarditis ). You have received a new heart valve ...

  14. On-line liquid phase micro-extraction based on drop-in-plug sequential injection lab-at-valve platform for metal determination

    Energy Technology Data Exchange (ETDEWEB)

    Mitani, Constantina [Laboratory of Analytical Chemistry, Department of Chemistry, Aristotle University, Thessaloniki 54124 (Greece); Anthemidis, Aristidis N., E-mail: anthemid@chem.auth.gr [Laboratory of Analytical Chemistry, Department of Chemistry, Aristotle University, Thessaloniki 54124 (Greece)

    2013-04-10

    Highlights: ► Drop-in-plug micro-extraction based on SI-LAV platform for metal preconcentration. ► Automatic liquid phase micro-extraction coupled with FAAS. ► Organic solvents with density higher than water are used. ► Lead determination in environmental water and urine samples. -- Abstract: A novel automatic on-line liquid phase micro-extraction method based on drop-in-plug sequential injection lab-at-valve (LAV) platform was proposed for metal preconcentration and determination. A flow-through micro-extraction chamber mounted at the selection valve was adopted without the need of sophisticated lab-on-valve components. Coupled to flame atomic absorption spectrometry (FAAS), the potential of this lab-at-valve scheme is demonstrated for trace lead determination in environmental and biological water samples. A hydrophobic complex of lead with ammonium pyrrolidine dithiocarbamate (APDC) was formed on-line and subsequently extracted into an 80 μL plug of chloroform. The extraction procedure was performed by forming micro-droplets of aqueous phase into the plug of the extractant. All critical parameters that affect the efficiency of the system were studied and optimized. The proposed method offered good performance characteristics and high preconcentration ratios. For 10 mL sample consumption an enhancement factor of 125 was obtained. The detection limit was 1.8 μg L{sup −1} and the precision expressed as relative standard deviation (RSD) at 50.0 μg L{sup −1} of lead was 2.9%. The proposed method was evaluated by analyzing certified reference materials and applied for lead determination in natural waters and urine samples.

  15. Evaluation of the Main Steam Line Break Accident for the APR+ Standard Design using MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Park, M. H.; Kim, Y. S.; Hwang, Min Jeong; Sim, S. K. [Environment Energy Technology, Daejeon (Korea, Republic of); Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    As a part of licensing evaluation of the APR+ (Advanced Power Reactor +) standard design, Korea Institute of Nuclear Safety(KINS) performed safety evaluation of the APR+ Standard Safety Analysis Report(SSAR). The results of the safety evaluation of the APR+ Main Steam Line Break(MSLB) accident is presented for the most limiting post-trip return-to-power case with the single failure assumption of the Loss Of Offsite Power(LOOP). MARS-KS regulatory safety analysis code has been used to evaluate safety as well as the system behavior during MSLB accident. The MARS-KS analysis results are compared with the results of the MSLB safety analysis presented in the SSAR of the APR+. Independent safety evaluation has been performed using MARS-KS regulatory safety analysis code for the APR+ MSLB accident inside containment for the limiting case of the full power post-trip return-to-power. The results of MARS-KS analysis were compared with the results of the MSLB safety analysis presented in the APR+ SSAR. Due to higher cooldown of the MARS-KS analysis, the MARS-KS analysis results in a higher positive reactivity insertion into the core by the negative moderator and fuel temperature reactivity coefficients than the APR+ SSAR analysis. Both results show no return-to-power during the limiting case of the MSLB inside containment. However, APR+ SSAR moderator temperature reactivity insertion should be evaluated against the MARS-KS moderator density reactivity insertion for is conservatism. This study also clearly shows asymmetric thermal hydraulic behavior during the MSLB accident at intact and affected sides of the downcomer and the core. These asymmetric phenomena should be further investigated for the effects on the system design.

  16. Valve Disease

    Science.gov (United States)

    ... blood. There are 4 valves in the heart: tricuspid, pulmonary, mitral, and aortic. Two types of problems can disrupt blood flow through the valves: regurgitation or stenosis. Regurgitation is also called insufficiency or incompetence. Regurgitation happens when a valve doesn’ ...

  17. HDR-investigations of check valve closure and resultant water hammer effects

    International Nuclear Information System (INIS)

    Scholl, K.D.

    1983-01-01

    The presented investigations are based on the Loss of Coolant Accident (LOCA). They concentrate on the first blowdown phase after pipe break of a feedwater line. The effect of such a break is moderated by quick closing check valves, by which the loss of coolant water is reduced and optimal post accident conditions are obtained. Unfortunately the closure of the valve can cause high pressure peaks (water hammer effects) in the feedwater system which potentially could produce safety relevant secondary damage. The system loading by these effects has been analysed. The HDR-Investigation-results led to an improvement of the feedwater system safety by verifying damping measures of quick closing check valves. Pressure peaks obtained with undamped valves in the range of 300 bars, are reduced to zero or a few bars above the normal operation pressure in feedwater systems. For the analytical simulation of valve closure the following dominant acting forces are identified: the blowdown flow resistance of the valve cone and the damping pistong force. The analytical description and quantification of the forces depends on blowdown flow and valve friction parameters. These have been evaluated and are presented for practical use. (orig.)

  18. Valve for gas centrifuges

    Science.gov (United States)

    Hahs, Charles A.; Burbage, Charles H.

    1984-01-01

    The invention is a pneumatically operated valve assembly for simultaneously (1) closing gas-transfer lines connected to a gas centrifuge or the like and (2) establishing a recycle path between two of the lines so closed. The valve assembly is especially designed to be compact, fast-acting, reliable, and comparatively inexpensive. It provides large reductions in capital costs for gas-centrifuge cascades.

  19. Tight valve

    International Nuclear Information System (INIS)

    Guedj, F.

    1987-01-01

    This sealed valve is made with a valve seat, an axial valve with a rod fixed to its upper end, a thick bell surrounding the rod and welded by a thin join on the valve casing, a threated ring screwed onto the upper end of the rod and a magnet or electromagnet rotating the ring outside the bell [fr

  20. Development of a severe accident training simulator using a MELCOR code

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo; Jung, Won Dae

    2002-03-01

    Nuclear power plants' severe accidents are, despite of their rareness, very important in safety aspects, because of their huge damages when occurred. For the appropriate execution of severe accident strategy, more information for decision-making are required because of the uncertainties included in severe accidents. Earlier NRC raised concerns over severe accident training in the report NUREC/CR-477, and accordingly, developing effective training tools for severe accident were emphasized. In fact the training tools were requested from industrial area, nevertheless, few training tools were developed due to the uncertainties in severe accidents, lacks of analysis computer codes and technical limitations. SATS, the severe accident training simulator, is developed as a multi-purpose tools for severe accident training. SATS uses the calculation results of MELCOR, an integral severe accident analysis code, and with the help of SL-GMS graphic tools, provides dynamic displays of severe accident phenomena on the terminal of IBM PC. It aimed to have two main features: one is to provide graphic displays to represent severe accident phenomena and the other is to process and simulate severe accident strategy given by plant operators and TSC staffs. Severe accident strategies are basically composed of series of operations of available pumps, valves and other equipments. Whenever executing strategies with SATS, the trainee should follow the HyperKAMG, the on line version of the recently developed severe accident guidance (KAMG). Severe accident strategies are closely related to accidents scenarios. TLOFW and LOCA , two representative severe accident scenarios of Uljin 3,4, are developed as a built-in scenarios of SATS. Although SATS has some minor problems at this time, we expect SATS will be a good severe accident training tool after the appropriate addition of accident scenarios. Moreover, we also expect SATS will be a good advisory tool for the severe accident research

  1. On-line two-dimensional capillary electrophoresis with mass spectrometric detection using a fully electric isolated mechanical valve.

    Science.gov (United States)

    Kohl, Felix J; Montealegre, Cristina; Neusüß, Christian

    2016-04-01

    CE is becoming more and more important in many fields of bioanalytical chemistry. Besides optical detection, hyphenation to ESI-MS detection is increasingly applied for sensitive identification purposes. Unfortunately, many CE techniques and methods established in research and industry are not compatible to ESI-MS since essential components of the background electrolyte interfere in ES ionization. In order to identify unknown peaks in established CE methods, here, a heart-cut 2D-CE separation system is introduced using a fully isolated mechanical valve with an internal loop of only 20 nL. In this system, the sample is separated using potentially any non-ESI compatible method in the first separation dimension. Subsequently, the portion of interest is cut by the internal sample loop of the valve and reintroduced to the second dimension where the interfering compounds are removed, followed by ESI-MS detection. When comparing the separation efficiency of the system with the valve to a system using a continuous capillary only a slight increase in peak width is observed. Ultraviolet/visible detection is integrated in the first dimension for switching time determination, enabling reproducible cutting of peaks of interest. The feasibility of the system is successfully demonstrated by a 2D analysis of a BSA tryptic digest sample using a nonvolatile (phosphate based) background electrolyte in the first dimension. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  2. Check valve

    Science.gov (United States)

    Upton, H.A.; Garcia, P.

    1999-08-24

    A check valve for use in a GDCS of a nuclear reactor and having a motor driven disk including a rotatable armature for rotating the check valve disk over its entire range of motion is described. In one embodiment, the check valve includes a valve body having a coolant flow channel extending therethrough. The coolant flow channel includes an inlet end and an outlet end. A valve body seat is located on an inner surface of the valve body. The check valve further includes a disk assembly, sometimes referred to as the motor driven disc, having a counterweight and a disk shaped valve. The disk valve includes a disk base having a seat for seating with the valve body seat. The disk assembly further includes a first hinge pin member which extends at least partially through the disk assembly and is engaged to the disk. The disk valve is rotatable relative to the first hinge pin member. The check valve also includes a motor having a stator frame with a stator bore therein. An armature is rotatably positioned within the stator bore and the armature is coupled to the disk valve to cause the disk valve to rotate about its full range of motion. 5 figs.

  3. Check valve

    International Nuclear Information System (INIS)

    Upton, H.A.; Garcia, P.

    1999-01-01

    A check valve for use in a GDCS of a nuclear reactor and having a motor driven disk including a rotatable armature for rotating the check valve disk over its entire range of motion is described. In one embodiment, the check valve includes a valve body having a coolant flow channel extending therethrough. The coolant flow channel includes an inlet end and an outlet end. A valve body seat is located on an inner surface of the valve body. The check valve further includes a disk assembly, sometimes referred to as the motor driven disc, having a counterweight and a disk shaped valve. The disk valve includes a disk base having a seat for seating with the valve body seat. The disk assembly further includes a first hinge pin member which extends at least partially through the disk assembly and is engaged to the disk. The disk valve is rotatable relative to the first hinge pin member. The check valve also includes a motor having a stator frame with a stator bore therein. An armature is rotatably positioned within the stator bore and the armature is coupled to the disk valve to cause the disk valve to rotate about its full range of motion. 5 figs

  4. An Overall Investigation of Direct Vessel Injection Line Break Accidents of the ATLAS Facility

    International Nuclear Information System (INIS)

    Kim, Yeon-Sik; Choi, Ki-Yong; Cho, Seok; Kim, Bok-Deuk

    2015-01-01

    For parametric evaluations of direct vessel injection (DVI) line break scenarios, the pressurizer (PZR) pressure, core collapsed water level, and peak cladding temperature were investigated between the analyses and tests. The PZR pressure was mainly dependent upon the break flow model, e.g., discharge coefficient of the Henry-Fauske critical model. The core collapsed water level and peak cladding temperature were mainly dependent on the counter-current flow limit (CCFL) option of the fuel alignment plate (FAP). The CCFL option of the cross-over leg (COL) affected the PZR pressure owing to the loop seal clearings and seemed to have little effect on the core collapsed water level. Proper C d values and applicable CCFL options were summarized. C d values seemed to be dependent on the sizes of the DVI line break. The PZR pressure was mainly dependent on the break flow model, e.g., the discharge coefficient of the Henry-Fauske critical model. The core collapsed water level and peak cladding temperature were mainly dependent on the CCFL option of the FAP. The CCFL option of the COL affected the PZR pressure owing to loop seal clearings and seemed to have little effect on the core collapsed water level. From parametric evaluations, proper C d values and applicable CCFL options were suggested. The C d values seemed to be dependent on the sizes of the DVI line break. Although there was little difference in the CCFL options of the COL, the Ku-option was the preferred one for COLs' CCFL option. The CCFL options of the FAP appeared sensitive to the core collapsed water level and peak cladding temperature. The Ku-option of the FAP tended to negatively exaggerate the core behavior and showed excessively conservative results, especially on the peak cladding temperature. For smaller breaks, e.g., 25%, NA- and Wa-options would be applicable for the FAP. However, for larger breaks, e.g., 50%, the Wa-option of the FAP was the preferred one. Comparisons between the tests and

  5. Main steam line break accident simulation of APR1400 using the model of ATLAS facility

    Science.gov (United States)

    Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

    2018-02-01

    A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

  6. Control Valve

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Wayne R.

    2018-03-20

    A control valve includes a first conduit having a first inlet and a first outlet and defining a first passage; a second conduit having a second inlet and a second outlet and defining a second passage, the second conduit extending into the first passage such that the second inlet is located within the first passage; and a valve plate disposed pivotably within the first passage, the valve plate defining a valve plate surface. Pivoting of the valve plate within the first passage varies flow from the first inlet to the first outlet and the valve plate is pivotal between a first position and a second position such that in the first position the valve plate substantially prevents fluid communication between the first passage and the second passage and such that in the second position the valve plate permits fluid communication between the first passage and the second passage.

  7. Valve assembly

    International Nuclear Information System (INIS)

    Sandling, M.

    1981-01-01

    An improved valve assembly, used for controlling the flow of radioactive slurry, is described. Radioactive contamination of the air during removal or replacement of the valve is prevented by sucking air from the atmosphere through a portion of the structure above the valve housing. (U.K.)

  8. Thermal fatigue behavior of valves

    International Nuclear Information System (INIS)

    Moinereau, D.; Scliffet, L.; Capion, J.C.; Genette, P.

    1991-01-01

    This paper reports that valves of pressurized water reactors are exposed to thermal shocks during transient operations. The numerous thermal shock tests performed on valves on the EDF test facilities have shown the sensibility of fillets and geometrical discontinuities to thermal fatigue: cracks can appear in those areas and grow through the valve body. Valves systems designated as level 1 must be designed to withstand fatigue up to the second isolation valve: the relevant rule is specified in the paragraph B 3500 of the French RCCM code. It is a simplified method which doesn't require finite element calculations. Many valve systems have been designed according to this rule and have been operated without accident. However, in one case, important cracks were found in the fillet of a check-valve after numerous thermal shocks. Calculation of the valve's behavior according to the RCCM code to estimate the fatigue damage resulting from thermal shocks led to a low damage factor, which doesn't agree with the experimental results. This was confirmed by new testings and showed the inadequacy of B 3500 rule for thermal transients. On this base a new rule is proposed to estimate fatigue damage resulting from thermal shocks. An experimental program has been realized to validate this rule. Axisymetrical analytical mock-ups with different geometries and one check-valve in austenitic stainless steel 316 L have been submitted to hot thermal shocks of 210 degrees C magnitude

  9. RELAP5 simulation of surge line break accident using combined and best estimate plus uncertainty approaches

    International Nuclear Information System (INIS)

    Kristof, Marian; Kliment, Tomas; Petruzzi, Alessandro; Lipka, Jozef

    2009-01-01

    Licensing calculations in a majority of countries worldwide still rely on the application of combined approach using best estimate computer code without evaluation of the code models uncertainty and conservative assumptions on initial and boundary, availability of systems and components and additional conservative assumptions. However best estimate plus uncertainty (BEPU) approach representing the state-of-the-art in the area of safety analysis has a clear potential to replace currently used combined approach. There are several applications of BEPU approach in the area of licensing calculations, but some questions are discussed, namely from the regulatory point of view. In order to find a proper solution to these questions and to support the BEPU approach to become a standard approach for licensing calculations, a broad comparison of both approaches for various transients is necessary. Results of one of such comparisons on the example of the VVER-440/213 NPP pressurizer surge line break event are described in this paper. A Kv-scaled simulation based on PH4-SLB experiment from PMK-2 integral test facility applying its volume and power scaling factor is performed for qualitative assessment of the RELAP5 computer code calculation using the VVER-440/213 plant model. Existing hardware differences are identified and explained. The CIAU method is adopted for performing the uncertainty evaluation. Results using combined and BEPU approaches are in agreement with the experimental values in PMK-2 facility. Only minimal difference between combined and BEPU approached has been observed in the evaluation of the safety margins for the peak cladding temperature. Benefits of the CIAU uncertainty method are highlighted.

  10. Determination of trace metal ions via on-line separation and preconcentration by means of chelating Sepharose beads in a sequential injection lab-on-valve (SI-LOV) system coupled to electrothermal atomic absorption spectrometric detection

    DEFF Research Database (Denmark)

    Long, Xiangbao; Hansen, Elo Harald; Miró, Manuel

    2005-01-01

    The analytical performance of an on-line sequential injection lab-on-valve (SI-LOV) system using chelating Sepharose beads as sorbent material for the determination of ultra trace levels of Cd(II), Pb(II) and Ni(II) by electrothermal atomic absorption spectrometry (ETAAS) is described and discussed...

  11. Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

    International Nuclear Information System (INIS)

    Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

    2004-01-01

    The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

  12. Altitude valve for railway suspension control system

    Science.gov (United States)

    Zhang, Xuan; Zhang, Lihao; Li, Qingxuan; Chen, WanSong

    2017-09-01

    With the variation of people and material during vehicle service, the gravity of vehicle could be unbalanced. As a result it might cause accident. In order to solve this problem, altitude valve is assembled on board. It can adjust the gravity of vehicle by the intake and outlet progress of the spring in the altitude valve to prevent the tilt of vehicles.

  13. Avoidance of transmission line pressure oscillations in discrete hydraulic systems – by shaping of valve opening characteristics

    DEFF Research Database (Denmark)

    Hansen, Anders Hedegaard; Pedersen, Henrik C.; Bech, Michael Møller

    2015-01-01

    The architecture of multi pressure line discrete fluid power force systems imposes rapid pressure shifts in the actuator volumes. These fast shifts between pressure levels often introduce pressure oscillations in the actuator chamber and connecting pipes. The topic of this paper is to perform...... pressure shifts by changing the connection between various fixed pressure lines without introducing significant pressure oscillation. As a case study a discrete force system is utilised is a Power Take Off(PTO) system of a wave energy converter. Four pressure shifting algorithms are proposed...

  14. Assessing the efficiency of automatically controlled valves (ACV) for pipeline sectioning

    Energy Technology Data Exchange (ETDEWEB)

    Veiga, Leandro S. da; Silva, Marcos J.M. da; Leite, Joao Paulo de B.; Santos, Renata N.R. dos; Jardim, Rodrigo B.O.; Quinto, Thiago C. do [Petrobras Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil)

    2009-07-01

    In order to mitigate the effects caused by the rupture of a gas pipeline and following ASME B 31.8 recommendations, block valves are installed in these structures. However, many transportation companies also install devices capable of infer the occurrence of an accident in a gas pipeline. The most common devices are the ones that actuate when pressure in gas pipeline reaches a low value early established (PSL) and those which close valves due to high rate of pressure drop (line-break). Line-break has the function of identifying as fast as possible the occurrence of a rupture in a gas pipeline by high rate of pressure drop in that line. Although PSL presents a later actuation when compared to the line break, it represents redundancy to the line-break system, since it is able to isolate the segment where the accident happened even if other devices or the operator had not done it before. The growing of gas pipelines transport capacity has been generated transients capable of causing an erroneous shut down of the shut down valves (SDV). The aim of this paper, therefore, is to present how the operational limits of SDV can be overcome with remote operation using SCADA System. (author)

  15. Simulation and analysis of a main steam line transient with isolation valves closure and subsequent pipe break

    Energy Technology Data Exchange (ETDEWEB)

    Stevanovic, Vladimir; Studovic, Milovan [Faculty of Mechanical Engineering, University of Belgrade, Belgrade (Yugoslavia); Bratic, Aleksandar [Thermal Power Plant Nikola Tesla (Yugoslavia)

    1993-11-01

    Simulation and analysis of a real main steam line break transient at the coal fired 300 MW Drmno Thermal Power Plant have been performed by the computer code TEA-01. The methods and procedures used could be applied to a nuclear power plant. 9 refs., 6 figs.

  16. Safety of Ikata Nuclear Power Station from the accident of Three Mile Island

    International Nuclear Information System (INIS)

    Nonaka, Hiroshi

    1979-01-01

    The leak of radioactive substances occurred on March 28, 1979, in the No. 2 plant of Three Mile Island Nuclear Power Station, and this accident must be put to use to prevent similar accidents and to secure safety hereafter in the nuclear power stations being operated in Japan. In the TMI accident, too many problems concerning the operation management seemed to exist in a series of events. In this paper, a few matters related to the TMI accident among the aspects of the operation management in Ikata Nuclear Power Station are reported. As the problems of operation management, it is considered that the operation of the TMI plant was continued as the exit valve of auxiliary feed line was closed, that it took long time to close the root valve for a pressurizer relief valve manually, and that the ECCS was stopped manually. In TMI, the abnormal phenomenon of losing main feed water has occurred 6 times since the attainment of criticality in March, 1978, and the opening and sticking of pressurizer relief valves occurred at least twice in about 150 times of their actuation in the nuclear reactors designed by Babcock and Wilcox Co. In Ikata Nuclear Power Station, these problems are detected early and the suitable measures are taken immediately, therefore it never happens to continue the operation as the problems are left as they are. It is not conceivable that similar troubles occur many times. (Kako, I.)

  17. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  18. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  19. Mitral Valve Disease

    Science.gov (United States)

    ... for mitral valve replacement—mechanical valves (metal) or biological valves (tissue). The principal advantage of mechanical valves ... small risk of stroke due to blood clotting. Biological valves usually are made from animal tissue. Biological ...

  20. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  1. An analysis of Three Mile Island: the accident that shouldn't have happened

    International Nuclear Information System (INIS)

    Rubinstein, E.; Mason, J.F.

    1979-01-01

    The sequence of events in the nuclear reactor accident at Three Mile Island on March 28, 1979, is reported. Three problems thought to trigger the reactor accident were a persistent leak of reactor coolant, a closing of two valves in the auxillary feedwater system, and an apparent resin blockage in the transfer line that forced water back into the condensate lines of the air pumps. Hindsight indicates that a large amount of the damage to the reactor core could have been prevented if operators had closed the electromatic relief valve to end the loss of coolant and not throttled down the high pressure injection pumps in the emergency core cooling system. Steps taken to reestablish control of the reactor core are described

  2. Letter Report (ETN-98-0005) S-farm Overground Transfer (OGT) Line Design Comparison and BIO Evaluation

    International Nuclear Information System (INIS)

    HICKS, D.F.

    1999-01-01

    This document provides an evaluation of the detailed design for the 2414 Overground Transfer (OGT) line between S-Farm valve pits 241-S-B and 2414-0. The evaluation compares the design calculations to the design features, the important assumptions, and the required controls for TWRS BIO representative accident scenarios

  3. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  4. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  5. Quantitative assessment of an aortic and pulmonary valve function according to valve fenestration

    International Nuclear Information System (INIS)

    Mirkhani, S.H.; Golestani, M.G.; Hosini, M.; Kazemian, A.

    1999-01-01

    There are some reasons for malfunction of aortic and pulmonary valve like fibrosis, calcification, and atheroma. Although, in some papers fenestration were known as a pathologic sign, but it is not generally accepted, while this matter is important in choosing suitable Homograft Heart Valve. In this paper fenestrations and its size, numbers and situation effect was studied. We collected 98 hearts, the donors died because of accident, we excluded valves with atheroma, calcification, fibrosis and unequal cusps, 91 aortic and 93 pulmonary valves were given further consideration. We classified valves according to situation, number and size of fenestration. Each valve was tested with 104 cm of non-nal saline column pressure which is equal to 76 mm Hg. Valve efficacy was detected by fluid flow assay. With study of 184 valves, 95 had no fenestration, 64 had less than 2 fenestration and 25 had more than 2 fenestration. Valve efficacy in condition of less than 2 fenestration was more than others (p <0.01). Malfunction effects of fenestration increased in larger valve and it will be decreased if their situation would be marginal (free margin of cusp). In the comparison of aortic and pulmonary valve we saw that malfunction effect of fenestration in pulmonary valve was more than aortic valve. Our experience in Immam Khomeini Homograft Valve Bank has shown that a great deal of valves is fenestrated. It seems that fenestration must be considered as a quality criterion in homograft valve preparation, especially in pulmonary and large aortic valves; but complementary studies is necessary

  6. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  7. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    International Nuclear Information System (INIS)

    Lee, Seung Min; Sabundjian, Gaianê

    2017-01-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  8. A program package connected with a communication network for accident statistics of NPP, TPP, HPP and the power lines

    International Nuclear Information System (INIS)

    Madjarova, A.

    1993-01-01

    The package is designed for registration and analysis of accidents according to users' needs. A possibility is also provided for easy data transfer and access to data on implemented decisions. Special programmes are developed for NPP, TPP, HPP, electricity supply branch, regional distribution management and the National Electric Company. The system is open for local network connection and file exchange between the workstations. The dialogue features are user-friendly. The emergency situations are classified according to the requirements of the enacted in Bulgaria 'Regulations for Investigation, Classification and Recording of Accidents in Electric and Thermal Stations and Networks, 1993'. The unified data input provides a possibility for insertion of additional texts (remarks), correcting and updating. Data security tools are also envisaged. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

  10. Piezoelectric valve

    Science.gov (United States)

    Petrenko, Serhiy Fedorovich

    2013-01-15

    A motorized valve has a housing having an inlet and an outlet to be connected to a pipeline, a saddle connected with the housing, a turn plug having a rod, the turn plug cooperating with the saddle, and a drive for turning the valve body and formed as a piezoelectric drive, the piezoelectric drive including a piezoelectric generator of radially directed standing acoustic waves, which is connected with the housing and is connectable with a pulse current source, and a rotor operatively connected with the piezoelectric generator and kinematically connected with the rod of the turn plug so as to turn the turn plug when the rotor is actuated by the piezoelectric generator.

  11. Application of signature analysis for determining the operational readiness of motor-operated valves under blowdown test conditions

    International Nuclear Information System (INIS)

    Haynes, H.D.

    1990-01-01

    In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of motor-operated valves (MOVs). As part of this work, ORNL participated in the gate valve flow interrruption blowdown (GVFIB) tests carried out in Huntsville, Alabama. The tests provided an excellent opportunity to evaluate signature analysis methods for determining the operability of MOVs under accident conditions. ORNL acquired motor current and torque switch shaft angular position signatures on two test MOVs during several GVFIB tests. The reduction in operating ''margin'' of both MOVs due to the presence of additional valve running loads imposed by high flow was clearly observed in motor current and torque switch angular position signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing the signature analysis techniques. (orig.)

  12. Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Seok Cho; Park, Hyun Sik; Choi, Nam Hyun; Park, Yu Sun; Kim, Jong Rok; Bae, Byoung Uhn; Kim, Yeon Sik; Kim, Kyung Doo; Choi, Ki Yong; Song, Chul Hwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small

  13. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  14. Status of the Real-time On-line Decision Support (RODOS) system for off-site emergency management after nuclear and radiological accidents

    International Nuclear Information System (INIS)

    Raskov, W.; Ehrhardt, J.; Landman, C.; Pasler-Sauer, J.

    2006-01-01

    Under the auspices of its EURATOM Research Framework Programmes, the European Commission (EC) has supported the development of the comprehensive decision support system RODOS (Real-time On-line Decision Support) for off-site emergency management after nuclear accidents for more than a decade. Many national research programmes, research institutes and industrial collaborators contributed to the project, in particular the German Ministry of Environment, Nature Conservation and Reactor Safety (B MU). The RODOS system can be applied to accidental releases into the atmosphere and various aquatic environments within and across Europe. It provides coherent support before, during and after such a release to assist analysis of the situation and decision making about short and long-term countermeasures for mitigating the consequences with respect to health, the environment, and the economy. Appropriate interfaces exist with local and national radiological monitoring data systems, meteorological measurements and forecasts, and for the adaptation to local, regional and national conditions in Europe. Within the European Integrated Project EURANOS of the sixth Framework Programme, the RODOS system is being enhanced, among others, for radiological emergencies such as dirty bombs attacks, transport accidents and satellite crashes by extensions of the nuclide list, the source term characteristics and the atmospheric dispersion model

  15. Application of signature analysis for determining the operational readiness of motor-operated valves under blowdown test conditions

    International Nuclear Information System (INIS)

    Haynes, H.D.

    1988-01-01

    In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of Motor-Operated Valves (MOVs). As part of this work, ORNL participated in the Gate Valve Flow Interruption Blowdown (GVFIB) tests carried out in Huntsville, Alabama. The GVFIB tests were intended primarily to determine the behavior of motor-operated gate valves under the temperature, pressure, and flow conditions expected to be experienced by isolation valves in Boiling Water Reactors (BWRs) during a high energy line break (blowdown) outside of containment. In addition, the tests provided an excellent opportunity to evaluate signature analysis methods for determining the operational readiness of the MOVs under those accident conditions. ORNL acquired motor current and torque switch shaft angular position data on two test MOVs during various times of the GVFIB tests. The reduction in operating ''margin'' of both MOVs due to the presence of additional valve running loads imposed by high flow was clearly observed in motor current and torque switch angular position signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing signature analysis techniques. 1 ref.; 16 figs

  16. Postulated accidents

    International Nuclear Information System (INIS)

    Ullrich, W.

    1980-01-01

    This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

  17. Aortic valve bypass

    DEFF Research Database (Denmark)

    Lund, Jens T; Jensen, Maiken Brit; Arendrup, Henrik

    2013-01-01

    In aortic valve bypass (AVB) a valve-containing conduit is connecting the apex of the left ventricle to the descending aorta. Candidates are patients with symptomatic aortic valve stenosis rejected for conventional aortic valve replacement (AVR) or transcatheter aortic valve implantation (TAVI). ...

  18. On-line dynamic extraction and automated determination of readily bioavailable hexavalent chromium in solid substrates using micro-sequential injection bead-injection lab-on-valve hyphenated with electrothermal atomic absorption spectrometry

    DEFF Research Database (Denmark)

    Long, Xiangbao; Miró, Manuel; Hansen, Elo Harald

    2006-01-01

    A novel and miniaturized micro-sequential injection bead injection lab-on-valve (μSI-BI-LOV) fractionation system was developed for in-line microcolumn soil extraction under simulated environmental scenarios and accurate monitoring of the content of easily mobilisable hexavalent chromium in soil...... environments at the sub-low parts-per-million level. The flow system integrates dynamic leaching of hexavalent chromium using deionized water as recommended by the German Standard DIN 38414-S4 method; on-line pH adjustment of the extract by a 0.01 mol L-1 Tris-HNO3 buffer solution; isolation of the chromate...... polluted agricultural soil material (San Joaquin Soil-Baseline Trace Element Concentrations) with water-soluble Cr(VI) salts at different concentration levels. The potential of the μSI-BI-LOV set-up with renewable surfaces for flame-AAS determination of high levels of readily bioavailable chromate...

  19. Transcatheter aortic valve replacement

    Science.gov (United States)

    ... gov/ency/article/007684.htm Transcatheter aortic valve replacement To use the sharing features on this page, please enable JavaScript. Transcatheter aortic valve replacement (TAVR) is surgery to replace the aortic valve. ...

  20. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2006-07-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  1. A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

    International Nuclear Information System (INIS)

    Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

    2006-01-01

    This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

  2. Magnetically operated check valve

    Science.gov (United States)

    Morris, Brian G.; Bozeman, Richard J., Jr.

    1994-06-01

    A magnetically operated check valve is disclosed. The valve is comprised of a valve body and a movable poppet disposed therein. A magnet attracts the poppet to hold the valve shut until the force of fluid flow through the valve overcomes the magnetic attraction and moves the poppet to an unseated, open position. The poppet and magnet are configured and disposed to trap a magnetically attracted particulate and prevent it from flowing to a valve seating region.

  3. Steel-fabricated butterfly valves for condenser circulating water system

    International Nuclear Information System (INIS)

    Kawase, Hiroshi; Yasuoka, Masahiro; Nanao, Teruaki.

    1979-01-01

    The steel-fabricated butterfly valves, which are large in general, and gave rubber linings inside to prevent the corrosion due to sea Water, are utilized for the condenser circulating water systems of thermal and nuclear power plants. Cast iron butterfly valves, having been used hitherto, have some technical irrationalities, such as corrosion prevention, the techniques for manufacturing large castings, severe thermal transient operation. On the contrary, the steel plate-fabricated butterfly valves have the following advantages; much superior characteristics in strength, rigidity and shock resistance, the streamline shape of valve plates, the narrow width between two flanges, superior execution of works for rubber lining, the perfect sealed structure, safety to vibration, light weight and easy maintenance. The structural design and the main specifications for the steel plate butterfly valves with the nominal bore from 1350 mm to 3500 mm are presented. Concerning the design criteria, the torque of operating butterfly valves and the strength of valve bodies, valve plates and valve stems are explained. The performance tests utilizing the mock-up valve were carried out for the measurements of stress distribution, the deformation of valve body, the endurance and the operating torque. In the welding standards for steel plate butterfly valves, three kinds of welded parts are classified, and the inspection method for each part is stipulated. The vibration of the valves induced by flow vortexes and cavitation is explained. (Nakai, Y.)

  4. LINES

    Directory of Open Access Journals (Sweden)

    Minas Bakalchev

    2015-10-01

    Full Text Available The perception of elements in a system often creates their interdependence, interconditionality, and suppression. The lines from a basic geometrical element have become the model of a reductive world based on isolation according to certain criteria such as function, structure, and social organization. Their traces are experienced in the contemporary world as fragments or ruins of a system of domination of an assumed hierarchical unity. How can one release oneself from such dependence or determinism? How can the lines become less “systematic” and forms more autonomous, and less reductive? How is a form released from modernistic determinism on the new controversial ground? How can these elements or forms of representation become forms of action in the present complex world? In this paper, the meaning of lines through the ideas of Le Corbusier, Leonidov, Picasso, and Hitchcock is presented. Spatial research was made through a series of examples arising from the projects of the architectural studio “Residential Transformations”, which was a backbone for mapping the possibilities ranging from playfulness to exactness, as tactics of transformation in the different contexts of the contemporary world.

  5. What Is Heart Valve Surgery?

    Science.gov (United States)

    ... working correctly. Most valve replacements involve the aortic Tricuspid valve and mitral valves. The aortic valve separates ... where it shouldn’t. This is called incompetence, insufficiency or regurgitation. • Prolapse — mitral valve flaps don’t ...

  6. What Is Heart Valve Disease?

    Science.gov (United States)

    ... and replacing it with a man-made or biological valve. Biological valves are made from pig, cow, or human ... the valve. Man-made valves last longer than biological valves and usually don’t have to be ...

  7. Environmental qualification testing of TFE valve components

    International Nuclear Information System (INIS)

    Eyvindson, A.; Krasinski, W.; McCutcheon, R.

    1997-01-01

    Valves containing tetrafluoroethylene (TFE) components are being used in many CANDU Nuclear Generating Stations. However, some concerns remain about the performance of TFE after exposure to high levels of radiation. Stations must therefore ensure that such valves perform reliably after being exposed to postulated accident radiation dose levels. The current Ontario Hydro Environmental Qualification [EQ] program specifies much higher postulated radiation exposure than the original design, to account for conditions following a LOCA. Initial assessments indicated that Teflon components would require replacement. Proof of acceptable performance can remove the need for large scale replacement, avoiding a significant cost penalty and preserving benefits due to the superior performance of TFE-based seals. A test program was undertaken at Chalk River Laboratories (CRL) to investigate the performance of three valves after irradiation to 10 Mrad. Such valves are currently used at the Bruce B Nuclear Generating Station. Each contains TFE packing rings; one also has TFE seats. Two of the valves are used in the ECIS recovery system, while the third is used for instrumentation loop isolation or as drain valves. All are exposed to little or no radiation during normal use. Based on the results of the tests, all the valves tested will still meet functional and performance requirements after the TFE components have been exposed to 10 Mrad of irradiation. (author)

  8. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  9. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  10. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  11. Microfluidic sieve valves

    Science.gov (United States)

    Quake, Stephen R; Marcus, Joshua S; Hansen, Carl L

    2015-01-13

    Sieve valves for use in microfluidic device are provided. The valves are useful for impeding the flow of particles, such as chromatography beads or cells, in a microfluidic channel while allowing liquid solution to pass through the valve. The valves find particular use in making microfluidic chromatography modules.

  12. Rotary pneumatic valve

    Science.gov (United States)

    Hardee, Harry C.

    1991-01-01

    A rotary pneumatic valve which is thrust balanced and the pneumatic pressure developed produces only radial loads on the valve cylinder producing negligible resistance and thus minimal torque on the bearings of the valve. The valve is multiplexed such that at least two complete switching cycles occur for each revolution of the cylinder spindle.

  13. Mitral Valve Stenosis

    Science.gov (United States)

    ... the left ventricle from flowing backward. A defective heart valve fails to either open or close fully. Risk factors Mitral valve stenosis is less common today than it once was because the most common cause, ... other heart valve problems, mitral valve stenosis can strain your ...

  14. Aortic Valve Stenosis

    Science.gov (United States)

    ... most cases, doctors don't know why a heart valve fails to develop properly, so it isn't something you could have prevented. Calcium buildup on the valve. With age, heart valves may accumulate deposits of calcium (aortic valve ...

  15. Remote actuated valve implant

    Science.gov (United States)

    McKnight, Timothy E; Johnson, Anthony; Moise, Jr., Kenneth J; Ericson, Milton Nance; Baba, Justin S; Wilgen, John B; Evans, III, Boyd McCutchen

    2014-02-25

    Valve implant systems positionable within a flow passage, the systems having an inlet, an outlet, and a remotely activatable valve between the inlet and outlet, with the valves being operable to provide intermittent occlusion of the flow path. A remote field is applied to provide thermal or magnetic activation of the valves.

  16. Project installation the large equipment in line system in Brazil. Gas export line valve P-40 FPSO-MLS. Field Marlim Sul, Campos Basin, Brazil; Operacao de instalacao do maior equipamento no sistema in line ja realizado no Brasil. Valvula do gasoduto P-40 X FPSO-MLS. Campo de Marlim Sul, Bacia de Campos, Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Marcos Antonio Rodrigues; Fernandes, Paulo Tavares [PETROBRAS, Campos dos Goytacases, RJ (Brazil). Exploracao e Producao

    2005-07-01

    This work will approach the current level of development of the installation of connected underwater equipment to flexible lines in the underwater engineering operations in Campos' Basin. The project will show studies, analysis and simulations (through software developed by PETROBRAS) about the installation of the largest equipment laid in the 'in-line' system (connected to flexible lines) in Brazil - and one of the largest of the world: the ESDV (Emergency Shut Down Valve) of the gas pipeline P-40 x FPSO-MLS, in the South Marlim field, in Campos' Basin. This ESDV, of about 18.000 kg, 4 m height and 6,5 m length, has the purpose of assuring the safety conditions on the facilities, interrupting the gas flow exported for P-40 in case of emergency situations. Its installation opened a new alternative in releasing underwater equipment, using the ships that install the flexible lines. This operation occurred in June, 2004, and required the use of a second vessel for support and monitoring of the ESDV laying. The ESDV was installed at 400 m from FPSO-MLS, in a water depth of 1.137 m. This method shall be used broadly by the company in the implantation of the new units of Campos' Basin, and the upcoming studies must consider the gradual increase of the water depth in the new projects. This work will focus the technological development in this area, and one of its purposes is to foresee the future difficulties that can appear in the implantation of the production systems in deep and ultra-deep waters. (author)

  17. IE Information Notice No. 85-17, Supplement 1: Possible sticking of ASCO solenoid valves

    International Nuclear Information System (INIS)

    Jordan, E.L.

    1992-01-01

    This notice is to inform recipients of the results of follow up investigations regarding the reasons for sticking of Automatic Switch Company (ASCO) solenoid valves used to shut main steam isolation valves (MSIVs) under accident conditions. GE has recommend that the licensee replace the potentially contaminated MSIV solenoid valves and institute a periodic examination and cleaning of the MSIV solenoid valves. Grand Gulf has replaced the eight MSIV HTX832320V dual solenoid valves with fully environmentally qualified ASCO Model NP 8323A20E dual solenoid valves. The environmentally qualified valve Model NP 8323A20E was included in a control sample placed in the test ovens with the solenoid valves that stuck at Grand Gulf. The environmentally qualified model did not stick under the test conditions that cause sticking in the other solenoid valves

  18. Scissor thrust valve actuator

    Science.gov (United States)

    DeWall, Kevin G.; Watkins, John C; Nitzel, Michael E.

    2006-08-29

    Apparatus for actuating a valve includes a support frame and at least one valve driving linkage arm, one end of which is rotatably connected to a valve stem of the valve and the other end of which is rotatably connected to a screw block. A motor connected to the frame is operatively connected to a motor driven shaft which is in threaded screw driving relationship with the screw block. The motor rotates the motor driven shaft which drives translational movement of the screw block which drives rotatable movement of the valve driving linkage arm which drives translational movement of the valve stem. The valve actuator may further include a sensory control element disposed in operative relationship with the valve stem, the sensory control element being adapted to provide control over the position of the valve stem by at least sensing the travel and/or position of the valve stem.

  19. Three Mile Island accident

    International Nuclear Information System (INIS)

    Barre, B.; Olivier, E.; Roux, J.P.; Pelle, P.

    2010-01-01

    Deluded by equivocal instrumentation signals, operators at TMI-2 (Three Mile Island - unit 2) misunderstood what was going on in the reactor and for 2 hours were taking inadequate decisions that turned a reactor incident into a major nuclear event that led to the melting of about one third of the core. The TMI accident had worldwide impacts in the domain of nuclear safety. The main consequences in France were: 1) the introduction of the major accident approach and the reinforcement of crisis management; 2) the improvement of the reactor design, particularly that of the pressurizer valves; 3) the implementation of safety probabilistic studies; 4) a better taking into account of the feedback experience in reactor operations; and 5) a better taking into account of the humane factor in reactor safety. (A.C.)

  20. 3-D analysis of reactor loop isolation valves

    International Nuclear Information System (INIS)

    Dietrich, D.E.

    1975-01-01

    A full three-dimensional analysis for the design and operational loading conditions was performed on a 29 inch loop isolation valve using the Westinghouse finite element computer code. The 3-D analysis was employed for the valve design in place of utilizing the standard ASME valve design criteria. The valve design employs the design by analysis concept allowed for nuclear class valve. The valve design was evaluated for a set of independent load including pipe reactions and internal pressure. The design pipe reaction loads were based upon maximum fiber pipe stresses at yield for the bending moments, pipe membrane stresses at half yield for the axial load, and pipe maximum shear stress at half yield for the torsional moment. The valve design pressure was the system loop design pressure. The operating and accident condition evaluation included pipe reactions, extended structure forces, system pressure, and system thermal transients. The valve was analyzed for the normal operating, upset, emergency, and faulted loading conditions. These operating and accident conditions used various specified combinations of the supplied generic system pressure, deadweight, thermal, seismic, and LOCA pipe load components. The generic pipe loads are the worst possible postulated loads for any system design. These generic pipe load components were supplied as maximums and minimums so a simplified nozzle analysis was performed to determine the worst case combination for each loading condition. The valve design was shown to meet the design, operating, and accident condition requirements of the ASME code. The design by analysis concept for nuclear class 1 valves gave a significant reduction in required minimum wall thickness, 3.75 inches vs. 5.4 inches. These translate into significant material savings

  1. Oil pipeline valve automation for spill reduction

    Energy Technology Data Exchange (ETDEWEB)

    Mohitpour, Mo; Trefanenko, Bill [Enbridge Technology Inc, Calgary (Canada); Tolmasquim, Sueli Tiomno; Kossatz, Helmut [TRANSPETRO - PETROBRAS Transporte S.A., Rio de Janeiro, RJ (Brazil)

    2003-07-01

    Liquid pipeline codes generally stipulate placement of block valves along liquid transmission pipelines such as on each side of major river crossings where environmental hazards could cause or are foreseen to potentially cause serious consequences. Codes, however, do not stipulate any requirement for block valve spacing for low vapour pressure petroleum transportation, nor for remote pipeline valve operations to reduce spills. A review of pipeline codes for valve requirement and spill limitation in high consequence areas is thus presented along with a criteria for an acceptable spill volume that could be caused by pipeline leak/full rupture. A technique for deciding economically and technically effective pipeline block valve automation for remote operation to reduce oil spilled and control of hazards is also provided. In this review, industry practice is highlighted and application of the criteria for maximum permissible oil spill and the technique for deciding valve automation thus developed, as applied to ORSUB pipeline is presented. ORSUB is one of the three initially selected pipelines that have been studied. These pipelines represent about 14% of the total length of petroleum transmission lines operated by PETROBRAS Transporte S.A. (TRANSPETRO) in Brazil. Based on the implementation of valve motorization on these three pipeline, motorization of block valves for remote operation on the remaining pipelines is intended, depending on the success of these implementations, on historical records of failure and appropriate ranking. (author)

  2. How to insure quality valve remanufacture

    International Nuclear Information System (INIS)

    Scott, C.F.

    1991-01-01

    The importance of quality valve repair for the power generation industry is an obvious need for both the owner as well as the consumer. Whether valves are repaired in-line, on-site, or at a valve remanufacturing facility, the selection of a vendor is vital to meeting not only stringent quality requirements, but also to meet start-up schedules and budgets. In the past, the rule of thumb was that repair of a valve could cost approximately 50% of the cost of a new valve and still represent a significant savings to the end user. For power generation facilities, the fact that many valves are welded in not only makes repair more economical, but even vital to continuing normal operations. For those items not welded in, long lead times and higher prices for these normally exotic alloys make remanufactured valves even more attractive. However, even as these advantages of remanufacturing are obvious, some repair organizations continue to cut corners to meet profit demands. The result is suspect quality in some valves. This can lead to premature failures, possible reduced generating capacity, unscheduled outages, and even catastrophic results. Therefore, the choice of a repair organization must be made with care. As the author has said, repair is an obvious option, but the procurement should definitely involve more than just price comparisons. Evaluation must place the emphasis on quality and reliability. Several aspects should be thoroughly investigated and documented in the selection process. These include: personnel; equipment/facilities; procedures; and credentials

  3. Preventing accidents

    Science.gov (United States)

    2005-08-01

    As the most effective strategy for improving safety is to prevent accidents from occurring at all, the Volpe Center applies a broad range of research techniques and capabilities to determine causes and consequences of accidents and to identify, asses...

  4. Hardfacing and packings for improved valve performance

    International Nuclear Information System (INIS)

    Aikin, J.A.; Patrick, J.N.F.; Inglis, I.

    2003-01-01

    The CANDU Owners Group (COG), Chemistry, Materials and Components (CMC) Program has supported an ongoing program on valve maintenance and performance for several years. An overview is presented of recent work on iron-based hardfacing, packing qualification, friction testing of polytetrafluoroethylene (PTFE) packings, and an investigation of re-torquing valve packing. Based on this program, two new valve-packing materials have been qualified for use in CANDU stations. By doing this, CANDU maintenance can avoid having only one packing qualified for station use, as well as assess the potential impact of the industry trend towards using lower gland loads. The results from corrosion tests by AECL and the coefficient of friction studies at Battelle' s tribology testing facilities on Delcrome 910, an iron-based hardfacing alloy, indicate it is an acceptable replacement for Stellite 6 under certain conditions. This information can be used to update in-line valve purchasing specifications. The renewed interest in friction characteristics, and environmental qualification (EQ) of packing containing PTFE has resulted in a new test program in these areas. The COG-funded valve programs have resulted in modifications to design specifications for nuclear station in-line valves and have led to better maintenance practices and valve reliability. In the end, this means lower costs and cheaper electricity. (author)

  5. Development of linear flow rate control system for eccentric butter-fly valve

    International Nuclear Information System (INIS)

    Kwak, K. K.; Cho, S. W.; Park, J. S.; Cho, J. H.; Song, I. T.; Kim, J. G.; Kwon, S. J.; Kim, I. J.; Park, W. K.

    1999-12-01

    Butter-fly valves are advantageous over gate, globe, plug, and ball valves in a variety of installations, particularly in the large sizes. The purpose of this project development of linear flow rate control system for eccentric butter-fly valve (intelligent butter-fly valve system). The intelligent butter-fly valve system consist of a valve body, micro controller. The micro controller consist of torque control system, pressure censor, worm and worm gear and communication line etc. The characteristics of intelligent butter-fly valve system as follows: Linear flow rate control function. Digital remote control function. guard function. Self-checking function. (author)

  6. Which valve is which?

    Directory of Open Access Journals (Sweden)

    Pravin Saxena

    2015-01-01

    Full Text Available A 25-year-old man presented with a history of breathlessness for the past 2 years. He had a history of operation for Tetralogy of Fallot at the age of 5 years and history suggestive of Rheumatic fever at the age of 7 years. On echocardiographic examination, all his heart valves were severely regurgitating. Morphologically, all the valves were irreparable. The ejection fraction was 35%. He underwent quadruple valve replacement. The aortic and mitral valves were replaced by metallic valve and the tricuspid and pulmonary by tissue valve.

  7. Bioprosthetic Valve Fracture Improves the Hemodynamic Results of Valve-in-Valve Transcatheter Aortic Valve Replacement.

    Science.gov (United States)

    Chhatriwalla, Adnan K; Allen, Keith B; Saxon, John T; Cohen, David J; Aggarwal, Sanjeev; Hart, Anthony J; Baron, Suzanne J; Dvir, Danny; Borkon, A Michael

    2017-07-01

    Valve-in-valve (VIV) transcatheter aortic valve replacement (TAVR) may be less effective in small surgical valves because of patient/prosthesis mismatch. Bioprosthetic valve fracture (BVF) using a high-pressure balloon can be performed to facilitate VIV TAVR. We report data from 20 consecutive clinical cases in which BVF was successfully performed before or after VIV TAVR by inflation of a high-pressure balloon positioned across the valve ring during rapid ventricular pacing. Hemodynamic measurements and calculation of the valve effective orifice area were performed at baseline, immediately after VIV TAVR, and after BVF. BVF was successfully performed in 20 patients undergoing VIV TAVR with balloon-expandable (n=8) or self-expanding (n=12) transcatheter valves in Mitroflow, Carpentier-Edwards Perimount, Magna and Magna Ease, Biocor Epic and Biocor Epic Supra, and Mosaic surgical valves. Successful fracture was noted fluoroscopically when the waist of the balloon released and by a sudden drop in inflation pressure, often accompanied by an audible snap. BVF resulted in a reduction in the mean transvalvular gradient (from 20.5±7.4 to 6.7±3.7 mm Hg, P valve effective orifice area (from 1.0±0.4 to 1.8±0.6 cm 2 , P valves to facilitate VIV TAVR with either balloon-expandable or self-expanding transcatheter valves and results in reduced residual transvalvular gradients and increased valve effective orifice area. © 2017 American Heart Association, Inc.

  8. Accident management for severe accidents

    International Nuclear Information System (INIS)

    Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

    1988-01-01

    The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

  9. Mitral Valve Prolapse

    Science.gov (United States)

    ... valve syndrome . What happens during MVP? Watch an animation of mitral valve prolapse When the heart pumps ( ... our brochures Popular Articles 1 Understanding Blood Pressure Readings 2 Sodium and Salt 3 Heart Attack Symptoms ...

  10. Problem: Mitral Valve Regurgitation

    Science.gov (United States)

    ... each time the left ventricle contracts. Watch an animation of mitral valve regurgitation A leaking mitral valve ... Not Alone Popular Articles 1 Understanding Blood Pressure Readings 2 Sodium and Salt 3 Heart Attack Symptoms ...

  11. Problem: Heart Valve Regurgitation

    Science.gov (United States)

    ... should be completely closed For example: Watch an animation of mitral valve regurgitation A leaking mitral valve ... Not Alone Popular Articles 1 Understanding Blood Pressure Readings 2 Sodium and Salt 3 Heart Attack Symptoms ...

  12. Aortic valve surgery - open

    Science.gov (United States)

    ... gov/ency/article/007408.htm Aortic valve surgery - open To use the sharing features on this page, ... separates the heart and aorta. The aortic valve opens so blood can flow out. It then closes ...

  13. Corrosion of valve metals

    International Nuclear Information System (INIS)

    Draley, J.E.

    1976-01-01

    A general survey related to the corrosion of valve metals or film-forming metals. The way these metals corrode with some general examples is described. Valve metals form relatively perfect oxide films with little breakdown or leakage when anodized

  14. Mitral valve surgery - open

    Science.gov (United States)

    ... Taking warfarin (Coumadin) References Otto CM, Bonow RO. Valvular heart disease. In: Mann DL, Zipes DP, Libby P, Bonow ... A.M. Editorial team. Heart Surgery Read more Heart Valve Diseases Read more Mitral Valve Prolapse Read more A. ...

  15. Swing check valve

    International Nuclear Information System (INIS)

    Eminger, H.E.

    1977-01-01

    A swing check valve which includes a valve body having an inlet and outlet is described. A recess in the valve body designed to hold a seal ring and a check valve disc swingable between open and closed positions. The disc is supported by a high strength wire secured at one end in a support spacer pinned through bearing blocks fixed to the valve body and at its other end in a groove formed on the outer peripheral surface of the disc. The parts are designed and chosen such to provide a lightweight valve disc which is held open by minimum velocity of fluid flowing through the valve which thus reduces oscillations and accompanying wear of bearings supporting the valve operating parts. (Auth.)

  16. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  17. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    International Nuclear Information System (INIS)

    Fuller, R.; Harrell, J.

    1996-01-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves

  18. Early results of gate valve flow interruption blowdown tests

    International Nuclear Information System (INIS)

    DeWall, K.G.

    1988-01-01

    The preliminary results of the USNRC/INEL high-energy BWR line break flow interruption testing are presented. Two representative nuclear valve assemblies were cycled under design basis Reactor Water Cleanup pipe break conditions to provide input for the technical basis for resolving the Nuclear Regulatory Commission's Generic Issue 87. The effects of the blowdown hydraulic loadings on valve operability, especially valve closure stem forces, were studied. The blowdown tests showed that, given enough thrust, typical gate valves will close against the high flow resulting from a line break. The tests also showed that proper operator sizing depends on the correct identification of values for the sizing equation. Evidence exists that values used in the past may not be conservative for all valve applications. The tests showed that improper operator lock ring installation following test or maintenance can invalidate in-situ test results and prevent the valve from performing its design function. 2 refs., 12 figs., 2 tabs

  19. Mitral Valve Prolapse

    Science.gov (United States)

    Mitral valve prolapse (MVP) occurs when one of your heart's valves doesn't work properly. The flaps of the valve are "floppy" and ... to run in families. Most of the time, MVP doesn't cause any problems. Rarely, blood can ...

  20. Overflow control valve

    International Nuclear Information System (INIS)

    Kessinger, B.A.; Hundal, R.; Parlak, E.A.

    1982-01-01

    An overflow control valve for use in a liquid sodium coolant pump tank which can be remotely engaged with and disengaged from the pump tank wall to thereby permit valve removal. An actuating shaft for controlling the valve also has means for operating a sliding cylinder against a spring to retract the cylinder from sealing contact with the pump tank nozzle. (author)

  1. Fluid control valves

    International Nuclear Information System (INIS)

    Rankin, J.

    1980-01-01

    A fluid control valve is described in which it is not necessary to insert a hand or a tool into the housing to remove the valve seat. Such a valve is particularly suitable for the control of radioactive fluids since maintenance by remote control is possible. (UK)

  2. A remote control valve

    International Nuclear Information System (INIS)

    Cachard, Maurice de; Dumont, Maurice.

    1976-01-01

    This invention concerns a remote control valve for shutting off or distributing a fluid flowing at a high rate and low pressure. Among the different valves at present in use, electric valves are the most recommended for remote control but their reliability is uncertain and they soon become costly when large diameter valves are used. The valve described in this invention does away with this drawback owing to its simplicity and the small number of moving parts, this makes it particularly reliable. It mainly includes: a tubular body fitted with at least one side opening; at least one valve wedge for this opening, coaxial with the body, and mobile; a mobile piston integral with this wedge. Several valves to the specifications of this invention can be fitted in series (a shut-off valve can be used in conjunction with one or more distribution valves). The fitting and maintenance of the valve is very simple owing to its design. It can be fabricated in any material such as metals, alloys, plastics and concrete. The structure of the valve prevents the flowing fluid from coming into contact with the outside environment, thereby making it particularly suitable in the handling of dangerous or corrosive fluids. Finally, the opening and shutting of the valve occurs slowly, thereby doing away with the water hammer effect so frequent in large bore pipes [fr

  3. Heart Valve Diseases

    Science.gov (United States)

    Your heart has four valves. Normally, these valves open to let blood flow through or out of your heart, and then shut to keep it from flowing ... close tightly. It's one of the most common heart valve conditions. Sometimes it causes regurgitation. Stenosis - when ...

  4. Motor operated valves problems tests and simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pinier, D.; Haas, J.L.

    1996-12-01

    An analysis of the two refusals of operation of the EAS recirculation shutoff valves enabled two distinct problems to be identified on the motorized valves: the calculation methods for the operating torques of valves in use in the power plants are not conservative enough, which results in the misadjustement of the torque limiters installed on their motorizations, the second problem concerns the pressure locking phenomenon: a number of valves may entrap a pressure exceeding the in-line pressure between the disks, which may cause a jamming of the valve. EDF has made the following approach to settle the first problem: determination of the friction coefficients and the efficiency of the valve and its actuator through general and specific tests and models, definition of a new calculation method. In order to solve the second problem, EDF has made the following operations: identification of the valves whose technology enables the pressure to be entrapped: the tests and numerical simulations carried out in the Research and Development Division confirm the possibility of a {open_quotes}boiler{close_quotes} effect: determination of the necessary modifications: development and testing of anti-boiler effect systems.

  5. Motor operated valves problems tests and simulations

    International Nuclear Information System (INIS)

    Pinier, D.; Haas, J.L.

    1996-01-01

    An analysis of the two refusals of operation of the EAS recirculation shutoff valves enabled two distinct problems to be identified on the motorized valves: the calculation methods for the operating torques of valves in use in the power plants are not conservative enough, which results in the misadjustement of the torque limiters installed on their motorizations, the second problem concerns the pressure locking phenomenon: a number of valves may entrap a pressure exceeding the in-line pressure between the disks, which may cause a jamming of the valve. EDF has made the following approach to settle the first problem: determination of the friction coefficients and the efficiency of the valve and its actuator through general and specific tests and models, definition of a new calculation method. In order to solve the second problem, EDF has made the following operations: identification of the valves whose technology enables the pressure to be entrapped: the tests and numerical simulations carried out in the Research and Development Division confirm the possibility of a open-quotes boilerclose quotes effect: determination of the necessary modifications: development and testing of anti-boiler effect systems

  6. Note nuclear accidents combat

    International Nuclear Information System (INIS)

    1989-01-01

    In this document the starting points are described which underlie the new framework for the nuclear-accident combat in the Netherlands. All the elaboration of this is indicated in main lines. The juridical consequences of the proposed structure are enlightened and the sequel activities are indicated. (H.W.). 6 figs.; 8 tabs

  7. Heavy gas valves

    Energy Technology Data Exchange (ETDEWEB)

    Steier, L [Vereinigte Armaturen Gesellschaft m.b.H., Mannheim (Germany, F.R.)

    1979-01-01

    Heavy gas valves must comply with special requirements. Apart from absolute safety in operation there are stringent requirements for material, sealing and ease of operation even in the most difficult conditions. Ball valves and single plate pipe gate valves lateral sealing rings have a dual, double sided sealing effect according to the GROVE sealing system. Single plate gate valves with lateral protective plates are suitable preferably for highly contaminated media. Soft sealing gate valves made of cast iron are used for low pressure applications.

  8. Nuclear accidents

    International Nuclear Information System (INIS)

    1987-01-01

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  9. The use of valves in the SAGD process

    Energy Technology Data Exchange (ETDEWEB)

    Romano, Michael A. [Global Marketing, Oil and Gas, Tyco Valves and Controls (United States)

    2011-07-01

    Steam-assisted gravity drainage (SAGD) is a developing technology, the aim of which is to increase production of bitumen while minimizing its environmental footprint. Valves must meet the process conditions of the operations, which depend on weel depth: deeper reservoirs of bitumen require higher steam injection pressure. A wide range of valves is used throughout the SAGD process. In the water softening plant, butterfly and process lined valves are used. HP gate valves are used for isolation, globe valves for vents/drains/bypasses, along with ARC valves for steam and booster pump projection with steam traps on injection lines in steam injection. Isolation valves are used throughout the low pressure process including ball, gate and triple-offset valves. Pressure management is carried out on all pressure vessels and lines. Control and choke valves are installed on well pads and production. Instrumentation, actuation and controls are installed throughout. In the ideal situation, suppliers and process engineers would work together in the early stages of a project.

  10. Relief valve testing study

    International Nuclear Information System (INIS)

    BROMM, R.D.

    2001-01-01

    Reclosing pressure-actuated valves, commonly called relief valves, are designed to relieve system pressure once it reaches the set point of the valve. They generally operate either proportional to the differential between their set pressure and the system pressure (gradual lift) or by rapidly opening fully when the set pressure is reached (pop action). A pop action valve allows the maximum fluid flow through the valve when the set pressure is reached. A gradual lift valve allows fluid flow in proportion to how much the system pressure has exceeded the set pressure of the valve (in the case of pressure relief) or has decreased below the set pressure (vacuum relief). These valves are used to protect systems from over and under pressurization. They are used on boilers, pressure vessels, piping systems and vacuum systems to prevent catastrophic failures of these systems, which can happen if they are under or over pressurized beyond the material tolerances. The construction of these valves ranges from extreme precision of less than a psi tolerance and a very short lifetime to extremely robust construction such as those used on historic railroad steam engines that are designed operate many times a day without changing their set pressure when the engines are operating. Relief valves can be designed to be immune to the effects of back pressure or to be vulnerable to it. Which type of valve to use depends upon the design requirements of the system

  11. Glovebox pressure relief and check valve

    International Nuclear Information System (INIS)

    Blaedel, K.L.

    1986-01-01

    This device is a combined pressure relief valve and check valve providing overpressure protection and preventing back flow into an inert atmosphere enclosure. The pressure relief is embodied by a submerged vent line in a mercury reservior, the releif pressure being a function of the submerged depth. The pressure relief can be vented into an exhaust system and the relieving pressure is only slightly influenced by the varying pressure in the exhaust system. The check valve is embodied by a ball which floats on the mercury column and contacts a seat whenever vacuum exists within the glovebox enclosure. Alternatively, the check valve is embodied by a vertical column of mercury, the maximum back pressure being a function of the height of the column of mercury

  12. Glovebox pressure relief and check valve

    Energy Technology Data Exchange (ETDEWEB)

    Blaedel, K.L.

    1986-03-17

    This device is a combined pressure relief valve and check valve providing overpressure protection and preventing back flow into an inert atmosphere enclosure. The pressure relief is embodied by a submerged vent line in a mercury reservior, the releif pressure being a function of the submerged depth. The pressure relief can be vented into an exhaust system and the relieving pressure is only slightly influenced by the varying pressure in the exhaust system. The check valve is embodied by a ball which floats on the mercury column and contacts a seat whenever vacuum exists within the glovebox enclosure. Alternatively, the check valve is embodied by a vertical column of mercury, the maximum back pressure being a function of the height of the column of mercury.

  13. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

  14. Preliminary observations of gate valve flow interruption tests, Phase 2

    International Nuclear Information System (INIS)

    Steele, R. Jr.; DeWall, K.G.

    1990-01-01

    This paper presents preliminary observations from the US Nuclear Regulatory Commission/Idaho National Engineering Laboratory Flexible Wedge Gate Valve Qualification and Flow Interruption Test Program, Phase 2. The program investigated the ability of selected boiling water reactor (BWR) process line valves to perform their containment isolation function at high energy pipe break conditions and other more normal flow conditions. The fluid and valve operating responses were measured to provide information concerning valve and operator performance at various valve loadings so that the information could be used to assess typical nuclear industry motor operator sizing equations. Six valves were tested, three 6-in. isolation valves representative of those used in reactor water cleanup systems in BWRs and three 10-in. isolation valves representative of those used in BWR high pressure coolant injection (HPCI) steam lines. The concern with these normally open isolation valves is whether they will close in the event of a downstream pipe break outside of containment. The results of this testing will provide part of the technical insights for NRC efforts regarding Generic Issue 87 (GI-87), Failure of the HPCI Steam Line Without Isolation, which includes concerns about the uncertainties in gate valve motor operator sizing and torque switch settings for these BWR containment isolation valves. As of this writing, the Phase 2 test program has just been completed. Preliminary observations made in the field confirmed most of the results from the Phase 1 test program. All six valves closing in high energy water, high energy steam, and high pressure cold water require more force to close than would be calculated using the typical variables in the standard industry motor operator sizing equations

  15. Sequential injection-bead injection-lab-on-valve schemes for on-line solid phase extraction and preconcentration of ultra-trace levels of heavy metals with determination by electrothermal atomic absorption spectrometry and inductively coupled plasma mass spectrometry

    International Nuclear Information System (INIS)

    Wang Jianhua; Hansen, Elo Harald; Miro, Manuel

    2003-01-01

    This communication presents an overview of the state-of-the-art of the exploitation of sequential injection (SI)-bead injection (BI)-lab-on-valve (LOV) schemes for automatic on-line sample pre-treatments interfaced with ETAAS and ICPMS detection as conducted in the authors' group. The discussions are focused on the applications of SI-BI-LOV protocols for on-line microcolumn based solid phase extraction of ultra-trace levels of heavy metals, employing the so-called renewable surface separation and preconcentration manipulatory scheme. Two types of sorbents have been employed as packing material, that is, the hydrophilic SP Sephadex C-25 cation exchange and iminodiacetate based Muromac A-1 chelating resins, and the hydrophobic poly(tetrafluoroethylene) (PTFE) and poly(styrene-divinylbenzene) copolymer alkylated with octadecyl groups (C 18 -PS/DVB). Using ETAAS as detection device, the easy-to-handle hydrophilic renewable reactors hold the features of improved R.S.D.s and LODs as compared to those operated in the conventional, permanent mode, in addition to the elimination of flow resistance. The hydrophobic columns fall into two categories, that is, the renewable one packed with C 18 -PS/DVB beads entails analogous R.S.D.s and LODs with respect to the conventional approach, while those with PTFE beads result in slightly inferior R.S.D.s and LODs by similar comparison, yet offering a wider dynamic range than when using an external permanent column. Moreover, the hydrophilic materials result in much higher enrichment of the analyte than the hydrophobic ones, although PTFE is the packing material that exhibits the best retention efficiency

  16. Guide to prosthetic cardiac valves

    International Nuclear Information System (INIS)

    Morse, D.; Steiner, R.M.; Fernandez, J.

    1985-01-01

    This book contains 10 chapters. Some of the chapter titles are: The development of artificial heart valves: Introduction and historical perspective; The radiology of prosthetic heart valves; The evaluation of patients for prosthetic valve implantation; Pathology of cardiac valve replacement; and Bioengineering of mechanical and biological heart valve substitutes

  17. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    Dallman, R.J.; Gottula, R.C.; Holcomb, E.E.; Jouse, W.C.; Wagoner, S.R.; Wheatley, P.D.

    1987-05-01

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

  18. Intelligent Flow Control Valve

    Science.gov (United States)

    Kelley, Anthony R (Inventor)

    2015-01-01

    The present invention is an intelligent flow control valve which may be inserted into the flow coming out of a pipe and activated to provide a method to stop, measure, and meter flow coming from the open or possibly broken pipe. The intelligent flow control valve may be used to stop the flow while repairs are made. Once repairs have been made, the valve may be removed or used as a control valve to meter the amount of flow from inside the pipe. With the addition of instrumentation, the valve may also be used as a variable area flow meter and flow controller programmed based upon flowing conditions. With robotic additions, the valve may be configured to crawl into a desired pipe location, anchor itself, and activate flow control or metering remotely.

  19. Cryogenic Cam Butterfly Valve

    Science.gov (United States)

    McCormack, Kenneth J. (Inventor)

    2016-01-01

    A cryogenic cam butterfly valve has a body that includes an axially extending fluid conduit formed there through. A disc lug is connected to a back side of a valve disc and has a circular bore that receives and is larger than a cam of a cam shaft. The valve disc is rotatable for a quarter turn within the body about a lug axis that is offset from the shaft axis. Actuating the cam shaft in the closing rotational direction first causes the camming side of the cam of the cam shaft to rotate the disc lug and the valve disc a quarter turn from the open position to the closed position. Further actuating causes the camming side of the cam shaft to translate the valve disc into sealed contact with the valve seat. Opening rotational direction of the cam shaft reverses these motions.

  20. Low noise control valve

    International Nuclear Information System (INIS)

    Christie, R.S.

    1975-01-01

    Noise is one of the problems associated with the use of any type of control valve in systems involving the flow of fluids. The advent of OSHA standards has prompted control valve manufacturers to design valves with special trim to lower the sound pressure level to meet these standards. However, these levels are in some cases too high, particularly when a valve must be located in or near an area where people are working at tasks requiring a high degree of concentration. Such locations are found around and near research devices and in laboratory-office areas. This paper describes a type of fluid control device presently being used at PPL as a bypass control valve in deionized water systems and designed to reduce sound pressure levels considerably below OSHA standards. Details of the design and construction of this constant pressure drop variable flow control valve are contained in the text and are shown in photographs and drawings. Test data taken are included

  1. Durability Tests of Ball Valve Prototype with Flowmeter Operation

    Science.gov (United States)

    Rogula, J.; Romanik, G.

    2018-02-01

    The results of the investigation of the prototypical ball valve are presented in this article. The innovation of the tested valve is a ball with a built-in measuring orifice. The valve has been subjected to durability tests. Leakage under three temperatures: ambient, -30°C and +100°C was analyzed. Sealing elements of the valve were tested for roughness and deviation of shape before and after the cycles of operation. Ball valve operation means cycles of open/close. It was planned to perform 1000 cycles at each temperature condition accordingly. Tests of the valve were performed under gas pressure equal to 10 MPa. The research was carried out under the Operational Program "Intelligent Development" (POIR 01.01.01-00-0013 / 15 "Development of devices for measurement of media flow on industrial trunk-lines".

  2. Improvements to the RELAP/SCDAPSIM of Laguna Verde model for the analysis of transients and accidents; Mejoras al modelo de Laguna Verde de RELAP/SCDAPSIM para el analisis de transitorios y accidentes

    Energy Technology Data Exchange (ETDEWEB)

    Amador G, R.; Castillo D, R.; Ortiz V, J.; Araiza M, E.; Martinez C, E., E-mail: rogelio.castillo@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    This work presents the improvements to the integral model of the nuclear power plant of Laguna Verde for the RELAP/SCDAPSIM code, for the simulation of transients and severe accidents. The model includes a new detailed geometry of the steam lines, as well as improvements in the performance of the emergency systems. A primary containment model has also been created, which will be used to analyze the effect of safety valve and relief valve discharges to the wet well suppression pool and the effect of the rupture of a recirculation loop on the dry well. The simulations performed with the new model show that the changes made improve the prediction of the phenomenology involved during transients and accidents. (Author)

  3. Aortic or Mitral Valve Replacement With the Biocor and Biocor Supra

    Science.gov (United States)

    2017-04-26

    Aortic Valve Insufficiency; Aortic Valve Regurgitation; Aortic Valve Stenosis; Aortic Valve Incompetence; Mitral Valve Insufficiency; Mitral Valve Regurgitation; Mitral Valve Stenosis; Mitral Valve Incompetence

  4. Proving test on the reliability for nuclear valves

    International Nuclear Information System (INIS)

    Kajiyama, Yasuo; Tashiro, Hisao; Uga, Takeo; Maeda, Shunichi.

    1986-01-01

    Since valves are the most common components, they could be the most frequent causes of troubles in nuclear power plants. This proving test, therefore, has an important meaning to examine and verify the reliability of various valves under simulating conditions of abnormal and transient operations of the nuclear power plant. The test was performed mainly for the various types and pressure ratings of valves which were used in the primary and secondary systems in BWR and PWR nuclear power plants and which had major operating or safety related functions in those nuclear power plants. The results of the proving test, confirmed for more than four years, showed relatively favourable performance of the tested valves. It is concluded that performances of valves including operability, seat sealing and structural integrity were proved under the thermal cycling, vibration and pipe reaction load conditions. Operating functions during and after accident such as loss of coolant accident were satisfactory. From these results, it was considered that the purpose of this proving test was satisfactorily fulfilled. Several data accumulated by the test would be useful to get better reliability if it was evaluated with the actually experienced data of valves in the nuclear power plants. (Nogami, K.)

  5. Magnetic Check Valve

    Science.gov (United States)

    Morris, Brian G.; Bozeman, Richard J., Jr.

    1994-01-01

    Poppet in proposed check valve restored to closed condition by magnetic attraction instead of spring force. Oscillations suppressed, with consequent reduction of wear. Stationary magnetic disk mounted just upstream of poppet, also containing magnet. Valve body nonmagnetic. Forward pressure or flow would push poppet away from stationary magnetic disk so fluid flows easily around poppet. Stop in valve body prevents poppet from being swept away. When flow stopped or started to reverse, magnetic attraction draws poppet back to disk. Poppet then engages floating O-ring, thereby closing valve and preventing reverse flow. Floating O-ring facilitates sealing at low loads.

  6. Butterfly valves for seawater

    International Nuclear Information System (INIS)

    Yamanaka, Katsuto

    1991-01-01

    Recently in thermal and nuclear power stations and chemical plants which have become large capacity, large quantity of cooling water is required, and mostly seawater is utilized. In these cooling water systems, considering thermal efficiency and economy, the pipings become complex, and various control functions are demanded. For the purpose, the installation of shut-off valves and control valves for pipings is necessary. The various types of valves have been employed, and in particular, butterfly valves have many merits in their function, size, structure, operation, maintenance, usable period, price and so on. The corrosion behavior of seawater is complicated due to the pollution of seawater, therefore, the environment of the valves used for seawater became severe. The structure and the features of the butterfly valves for seawater, the change of the structure of the butterfly valves for seawater and the checkup of the butterfly valves for seawater are reported. The corrosion of metallic materials is complicatedly different due to the locating condition of plants, the state of pipings and the condition of use. The corrosion countermeasures for butterfly valves must be examined from the synthetic viewpoints. (K.I.)

  7. Redo mitral valve surgery

    Directory of Open Access Journals (Sweden)

    Redoy Ranjan

    2018-03-01

    Full Text Available This study is based on the findings of a single surgeon’s practice of mitral valve replacement of 167 patients from April 2005 to June 2017 who developed symptomatic mitral restenosis after closed or open mitral commisurotomy. Both clinical and color doppler echocardiographic data of peri-operative and six months follow-up period were evaluated and compared to assess the early outcome of the redo mitral valve surgery. With male-female ratio of 1: 2.2 and after a duration of 6 to 22 years symptom free interval between the redo procedures, the selected patients with mitral valve restenosis undergone valve replacement with either mechanical valve in 62% cases and also tissue valve in 38% cases. Particular emphasis was given to separate the adhered pericardium from the heart completely to ameliorate base to apex and global contraction of the heart. Besides favorable post-operative clinical outcome, the echocardiographic findings were also encouraging as there was statistically significant increase in the mitral valve area and ejection fraction with significant decrease in the left atrial diameter, pressure gradient across the mitral valve and pulmonary artery systolic pressure. Therefore, in case of inevitable mitral restenosis after closed or open commisurotomy, mitral valve replacement is a promising treatment modality.

  8. Normal accidents

    International Nuclear Information System (INIS)

    Perrow, C.

    1989-01-01

    The author has chosen numerous concrete examples to illustrate the hazardousness inherent in high-risk technologies. Starting with the TMI reactor accident in 1979, he shows that it is not only the nuclear energy sector that bears the risk of 'normal accidents', but also quite a number of other technologies and industrial sectors, or research fields. The author refers to the petrochemical industry, shipping, air traffic, large dams, mining activities, and genetic engineering, showing that due to the complexity of the systems and their manifold, rapidly interacting processes, accidents happen that cannot be thoroughly calculated, and hence are unavoidable. (orig./HP) [de

  9. Diseases of the Tricuspid Valve

    Science.gov (United States)

    ... stenosis. Tricuspid Regurgitation Tricuspid regurgitation is also called tricuspid insufficiency or tricuspid incompetence. It means there is a ... require valve surgery. Tags: heart valves , tricuspid incompetence , ... tricuspid regurgitation , tricuspid stenosis , valve disease Related Links ...

  10. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  11. Radiation accidents

    International Nuclear Information System (INIS)

    Nenot, J.C.

    1996-01-01

    Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

  12. Experimental apparatus to test air trap valves

    Science.gov (United States)

    Lemos De Lucca, Y. de F.; de Aquino, G. A.; Filho, J. G. D.

    2010-08-01

    It is known that the presence of trapped air within water distribution pipes can lead to irregular operation or even damage to the distribution systems and their components. The presence of trapped air may occur while the pipes are being filled with water, or while the pumping systems are in operation. The formation of large air pockets can produce the water hammer phenomenon, the instability and the loss of pressure in the water distribution networks. As a result, it can overload the pumps, increase the consumption of electricity, and damage the pumping system. In order to avoid its formation, all of the trapped air should be removed through "air trap valves". In Brazil, manufacturers frequently have unreliable sizing charts, which cause malfunctioning of the "air trap valves". The result of these malfunctions causes accidents of substantial damage. The construction of a test facility will provide a foundation of technical information that will be used to help make decisions when designing a system of pipelines where "air trap valves" are used. To achieve this, all of the valve characteristics (geometric, mechanic, hydraulic and dynamic) should be determined. This paper aims to describe and analyze the experimental apparatus and test procedure to be used to test "air trap valves". The experimental apparatus and test facility will be located at the University of Campinas, Brazil at the College of Civil Engineering, Architecture, and Urbanism in the Hydraulics and Fluid Mechanics laboratory. The experimental apparatus will be comprised of various components (pumps, steel pipes, butterfly valves to control the discharge, flow meter and reservoirs) and instrumentation (pressure transducers, anemometer and proximity sensor). It should be emphasized that all theoretical and experimental procedures should be defined while taking into consideration flow parameters and fluid properties that influence the tests.

  13. Experimental apparatus to test air trap valves

    Energy Technology Data Exchange (ETDEWEB)

    Lemos De Lucca, Y de F [CTH-DAEE-USP/FAAP/UNICAMP (Brazil); Aquino, G A de [SABESP/UNICAMP (Brazil); Filho, J G D, E-mail: yvone.lucca@gmail.co [Water Resources Department, University of Campinas-UNICAMP, Av. Albert Einstein, 951, Cidade Universitaria-Barao Geraldo-Campinas, S.P., 13083-852 (Brazil)

    2010-08-15

    It is known that the presence of trapped air within water distribution pipes can lead to irregular operation or even damage to the distribution systems and their components. The presence of trapped air may occur while the pipes are being filled with water, or while the pumping systems are in operation. The formation of large air pockets can produce the water hammer phenomenon, the instability and the loss of pressure in the water distribution networks. As a result, it can overload the pumps, increase the consumption of electricity, and damage the pumping system. In order to avoid its formation, all of the trapped air should be removed through 'air trap valves'. In Brazil, manufacturers frequently have unreliable sizing charts, which cause malfunctioning of the 'air trap valves'. The result of these malfunctions causes accidents of substantial damage. The construction of a test facility will provide a foundation of technical information that will be used to help make decisions when designing a system of pipelines where 'air trap valves' are used. To achieve this, all of the valve characteristics (geometric, mechanic, hydraulic and dynamic) should be determined. This paper aims to describe and analyze the experimental apparatus and test procedure to be used to test 'air trap valves'. The experimental apparatus and test facility will be located at the University of Campinas, Brazil at the College of Civil Engineering, Architecture, and Urbanism in the Hydraulics and Fluid Mechanics laboratory. The experimental apparatus will be comprised of various components (pumps, steel pipes, butterfly valves to control the discharge, flow meter and reservoirs) and instrumentation (pressure transducers, anemometer and proximity sensor). It should be emphasized that all theoretical and experimental procedures should be defined while taking into consideration flow parameters and fluid properties that influence the tests.

  14. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  15. Radiation accidents

    International Nuclear Information System (INIS)

    Poplavskij, K.K.; Smorodintseva, G.I.

    1978-01-01

    On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

  16. Valve repair for traumatic tricuspid regurgitation.

    Science.gov (United States)

    Maisano, F; Lorusso, R; Sandrelli, L; Torracca, L; Coletti, G; La Canna, G; Alfieri, O

    1996-01-01

    The review of six cases of valve repair for traumatic tricuspid regurgitation in our institution and 74 in the literature in order to assess effective methods of treating this lesion. Tricuspid valve regurgitation is a rare complication of blunt chest trauma. Optimal treatment for this condition is still controversial ranging from long-term medical therapy to early surgical correction. We followed the cases of six consecutive patients with post-traumatic tricuspid incompetence who were successfully treated with reparative techniques. All patients were male and their ages ranged from 18 years to 42 years. Valve regurgitation was always secondary to blunt chest trauma due to motor vehicle accident. The mechanism of valve insufficiency was invariably anterior leaflet prolapse due to chordal or papillary muscle rupture associated with annular dilatation. Surgical procedures included Carpentier ring implant (5 patients), Bex posterior annuloplasty (1 patient), implant of artificial chordae (4 patients), papillary muscle reinsertion (2 patients), commissuroplasty (1 patient) and "artificial double orifice" technique (1 patient). Tricuspid insufficiency improved in all patients after the correction. No complications were recorded and all patients were asymptomatic at the follow-up. Since post-traumatic tricuspid regurgitation is effectively correctable with reparative techniques, early operation is recommended to relieve symptoms and to prevent right ventricular dysfunction.

  17. Danfos: Thermostatic Radiator Valves

    DEFF Research Database (Denmark)

    Gregersen, Niels; Oliver, James; Hjorth, Poul G.

    2000-01-01

    This problem deals with modelling the flow through a typical Danfoss thermostatic radiator valve.Danfoss is able to employ Computational Fluid Dynamics (CFD) in calculations of the capacity of valves, but an experienced engineer can often by rules of thumb "guess" the capacity, with a precision...

  18. Pedestrian injury risk functions based on contour lines of equal injury severity using real world pedestrian/passenger-car accident data.

    Science.gov (United States)

    Niebuhr, Tobias; Junge, Mirko; Achmus, Stefanie

    2013-01-01

    Injury risk assessment plays a pivotal role in the assessment of the effectiveness of Advanced Driver Assistance Systems (ADAS) as they specify the injury reduction potential of the system. The usual way to describe injury risks is by use of injury risk functions, i.e. specifying the probability of an injury of a given severity occurring at a specific technical accident severity (collision speed). A method for the generation of a family of risk functions for different levels of injury severity is developed. The injury severity levels are determined by use of a rescaled version of the Injury Severity Score (ISS) namely the ISSx. The injury risk curves for each collision speed is then obtained by fixing the boundary conditions and use of a case-by-case validated GIDAS subset of pedestrian-car accidents (N=852). The resultant functions are of exponential form as opposed to the frequently used logistic regression form. The exponential approach in combination with the critical speed value creates a new injury risk pattern better fitting for high speed/high energy crashes. Presented is a family of pedestrian injury risk functions for an arbitrary injury severity. Thus, the effectiveness of an ADAS can be assessed for mitigation of different injury severities using the same injury risk function and relying on the internal soundness of the risk function with regard to different injury severity levels. For the assessment of emergency braking ADAS, a Zone of Effective Endangerment Increase (ZEEI), the speed interval in which a one percent speed increase results at least in a one percent of injury risk increase, is defined. The methodology presented is kept in such general terms that a direct adaption to other accident configurations is easily done.

  19. Bioprinting a cardiac valve.

    Science.gov (United States)

    Jana, Soumen; Lerman, Amir

    2015-12-01

    Heart valve tissue engineering could be a possible solution for the limitations of mechanical and biological prostheses, which are commonly used for heart valve replacement. In tissue engineering, cells are seeded into a 3-dimensional platform, termed the scaffold, to make the engineered tissue construct. However, mimicking the mechanical and spatial heterogeneity of a heart valve structure in a fabricated scaffold with uniform cell distribution is daunting when approached conventionally. Bioprinting is an emerging technique that can produce biological products containing matrix and cells, together or separately with morphological, structural and mechanical diversity. This advance increases the possibility of fabricating the structure of a heart valve in vitro and using it as a functional tissue construct for implantation. This review describes the use of bioprinting technology in heart valve tissue engineering. Copyright © 2015 Elsevier Inc. All rights reserved.

  20. Space Vehicle Valve System

    Science.gov (United States)

    Kelley, Anthony R. (Inventor); Lindner, Jeffrey L. (Inventor)

    2014-01-01

    The present invention is a space vehicle valve system which controls the internal pressure of a space vehicle and the flow rate of purged gases at a given internal pressure and aperture site. A plurality of quasi-unique variable dimension peaked valve structures cover the purge apertures on a space vehicle. Interchangeable sheet guards configured to cover valve apertures on the peaked valve structure contain a pressure-activated surface on the inner surface. Sheet guards move outwardly from the peaked valve structure when in structural contact with a purge gas stream flowing through the apertures on the space vehicle. Changing the properties of the sheet guards changes the response of the sheet guards at a given internal pressure, providing control of the flow rate at a given aperture site.

  1. Multiple-port valve

    International Nuclear Information System (INIS)

    Doody, T.J.

    1978-01-01

    A multiple-port valve assembly is designed to direct flow from a primary conduit into any one of a plurality of secondary conduits as well as to direct a reverse flow. The valve includes two mating hemispherical sockets that rotatably receive a spherical valve plug. The valve plug is attached to the primary conduit and includes diverging passageways from that conduit to a plurality of ports. Each of the ports is alignable with one or more of a plurality of secondary conduits fitting into one of the hemispherical sockets. The other hemispherical socket includes a slot for the primary conduit such that the conduit's motion along that slot with rotation of the spherical plug about various axes will position the valve-plug ports in respect to the secondary conduits

  2. Analysis Of Feedwater Line Break Of APR1400 By MARS Code

    International Nuclear Information System (INIS)

    Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

    2011-01-01

    This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

  3. Evaluating the break flow for the 100% DVI line break accident of ATLAS using the RELAP5/MOD3.3 code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Suk Ho; You, Sung Chang; Kim, Han Gon [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2010-10-15

    An integral effect test database for major design basis accidents using the Advanced Test Loop for Accident Simulation (ATLAS) facility has been compiled by the Korea Atomic Energy Research Institute (KAERI). In order to effectively utilize the database, the Domestic Standard Problem (DSP) exercise was proposed and launched in 2009. As the first DSP exercise, scenario involving a 100% break of the DVI nozzle was determined by considering its technical importance including such phenomena as the break flow, loop seal clearing. The first DSP exercise was performed in an open calculation environment. Thus, integral effect test data were opened to the participants prior to code calculations. Ten domestic organizations including members of nuclear industry, a research institute, and universities participated in the DSP exercise using various best-estimate safety analysis codes and finally presented their code prediction results, comparing them to the experimental data. This paper presents the analysis results performed by NETEC as one of the first DSP exercise participants. This analysis focuses on the break flow phenomena and modeling

  4. Control valve sizing and specification: The first step

    International Nuclear Information System (INIS)

    Harkins, J.F.; Hoyle, E.D.

    1991-01-01

    Today's modern control valve can satisfy almost any application. Special trim, materials, operators, and body configurations have been developed to meet the most severe operating conditions. The missing link in the chain connecting design to application is often the interpretation and communication of the requirements for determining the proper valve for each application. This paper addresses an important but often neglected requirement for proper selection and sizing of control valves: the determination of correct input data. It presents criteria necessary to ensure that the data given the manufacturer accurately reflects the conditions under which the control valve will operate. It highlights the importance of communication between the system design engineer, the valve specifying engineer, and the control valve supplier, to ensure that the final system design meets the true requirements of the application. An example is provided of a simple liquid-handling system, for which line losses and variations in flow and equipment capacities are tabulated and requirements shown graphically on typical control valve characteristic curves. The effects of seemingly harmless, conservative assumptions regarding line losses, equipment capacities and selection, sizing practices, and the selection of various flow data can have on the final valve selection are illustrated. Also discussed is the proper selection of equipment and input data, based on the example

  5. Main feedwater valve diagnostics at Waterford 3 nuclear generating station

    International Nuclear Information System (INIS)

    Fitzgerald, W.V.

    1991-01-01

    Pneumatically-operated control valves are coming under increasing scrutiny in nuclear power plants because of their relatively high incident rate. The theory behind a device that could make performance evaluation of these valves simpler and more effective was first described at the original EPRI Power Plant Valve Symposium. The development of this Diagnostic System was completed in 1989, and it was recently used to troubleshoot two main feedwater valves at Louisiana Power and Light's Waterford 3 Power Station. During a cold snap last December, these valves failed to respond to the input signal and, as a result, the plant came off line. An incident report had to be filed, and the plant chose to contact the original equipment manufacturer (OEM) for assistance. This paper describes the original incident involving these valves and then gives a brief description of the diagnostic system and how it works. The balance of the paper then reviews how the OEM and plant personnel utilized the system to evaluate each component of the control valve assembly (I/P transducer, positioner, volume boosters, actuator, and valve body assembly). By simply stroking the valve and monitoring pneumatic signals and valve position, the problem was traced to a malfunctioning positioner and a volume booster that was leaking. The problems were corrected and new performance signatures run for the valves using the system to document their improved operation. This case study demonstrates how new Diagnostic Technology along with OEM involvement can effectively address problems with pneumatically-operated control valves so that root-cause solutions can be implemented

  6. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

    A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

  7. Gate valve performance prediction

    International Nuclear Information System (INIS)

    Harrison, D.H.; Damerell, P.S.; Wang, J.K.; Kalsi, M.S.; Wolfe, K.J.

    1994-01-01

    The Electric Power Research Institute is carrying out a program to improve the performance prediction methods for motor-operated valves. As part of this program, an analytical method to predict the stem thrust required to stroke a gate valve has been developed and has been assessed against data from gate valve tests. The method accounts for the loads applied to the disc by fluid flow and for the detailed mechanical interaction of the stem, disc, guides, and seats. To support development of the method, two separate-effects test programs were carried out. One test program determined friction coefficients for contacts between gate valve parts by using material specimens in controlled environments. The other test program investigated the interaction of the stem, disc, guides, and seat using a special fixture with full-sized gate valve parts. The method has been assessed against flow-loop and in-plant test data. These tests include valve sizes from 3 to 18 in. and cover a considerable range of flow, temperature, and differential pressure. Stem thrust predictions for the method bound measured results. In some cases, the bounding predictions are substantially higher than the stem loads required for valve operation, as a result of the bounding nature of the friction coefficients in the method

  8. Modeling valve leakage

    International Nuclear Information System (INIS)

    Bell, S.R.; Rohrscheib, R.

    1994-01-01

    The American Society of Mechanical Engineers (ASME) Code requires individual valve leakage testing for Category A valves. Although the U.S. Nuclear Regulatory Commission (USNRC) has recognized that it is more appropriate to test containment isolation valves in groups, as allowed by 10 CFR 50, Appendix J, a utility seeking relief from these Code requirements must provide technical justification for the relief and establish a conservative alternate acceptance criteria. In order to provide technical justification for group testing of containment isolation valves, Illinois Power developed a calculation (model) for determining the size of a leakage pathway in a valve disc or seat for a given leakage rate. The model was verified experimentally by machining leakage pathways of known size and then measuring the leakage and comparing this value to the calculated value. For the range of values typical of leakage rate testing, the correlation between the experimental values and calculated values was quote good. Based upon these results, Illinois Power established a conservative acceptance criteria for all valves in the inservice testing (IST) program and was granted relief by the USNRC from the individual leakage testing requirements of the ASME Code. This paper presents the results of Illinois Power's work in the area of valve leakage rate testing

  9. Face-Sealing Butterfly Valve

    Science.gov (United States)

    Tervo, John N.

    1992-01-01

    Valve plate made to translate as well as rotate. Valve opened and closed by turning shaft and lever. Interactions among lever, spring, valve plate, and face seal cause plate to undergo combination of translation and rotation so valve plate clears seal during parts of opening and closing motions.

  10. GIANT PROSTHETIC VALVE THROMBUS

    Directory of Open Access Journals (Sweden)

    Prashanth Kumar

    2015-04-01

    Full Text Available Mechanical prosthetic valves are predisposed to bleeding, thrombosis & thromboembolic complications. Overall incidence of thromboembolic complications is 1% per year who are on oral anticoagulants, whereas bleeding complications incidence is 0.5% to 6.6% per year. 1, 2 Minimization of Scylla of thromboembolic & Charybdis of bleeding complication needs a balancing act of optimal antithrombotic therapy. We are reporting a case of middle aged male patient with prosthetic mitral valve presenting in heart failure. Patient had discontinued anticoagulants, as he had subdural hematoma in the past. He presented to our institute with a giant prosthetic valve thrombus.

  11. SAMEX: A severe accident management support expert

    International Nuclear Information System (INIS)

    Park, Soo-Yong; Ahn, Kwang-Il

    2010-01-01

    A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R and D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.

  12. Limit lines for risk

    International Nuclear Information System (INIS)

    Cox, D.C.; Baybutt, P.

    1982-01-01

    Approaches to the regulation of risk from technological systems, such as nuclear power plants or chemical process plants, in which potential accidents may result in a broad range of adverse consequences must take into account several different aspects of risk. These include overall or average risk, accidents posing high relative risks, the rate at which accident probability decreases with increasing accident consequences, and the impact of high frequency, low consequence accidents. A hypothetical complementary cumulative distribution function (CCDF), with appropriately chosen parametric form, meets all these requirements. The Farmer limit line, by contrast, places limits on the risks due to individual accident sequences, and cannot adequately account for overall risk. This reduces its usefulness as a regulatory tool. In practice, the CCDF is used in the Canadian nuclear licensing process, while the Farmer limit line approach, supplemented by separate qualitative limits on overall risk, is employed in the United Kingdom

  13. Valve monitoring ITI-MOVATS

    International Nuclear Information System (INIS)

    Moureau, S.

    1993-01-01

    ITI-MOVATS provides a wide range of test devices to monitor the performance of valves: motor operated gate or globe valve, butterfly valve, air operated valve, and check valve. The ITI-MOVATS testing equipment is used in the following three areas: actuator setup/baseline testing, periodic/post-maintenance testing, and differential pressure testing. The parameters typically measured with the MOVATS diagnostic system as well as the devices used to measure them are described. (Z.S.)

  14. Bioprosthetic Valve Fracture to Facilitate Transcatheter Valve-in-Valve Implantation.

    Science.gov (United States)

    Allen, Keith B; Chhatriwalla, Adnan K; Cohen, David J; Saxon, John T; Aggarwal, Sanjeev; Hart, Anthony; Baron, Suzanne; Davis, J Russell; Pak, Alex F; Dvir, Danny; Borkon, A Michael

    2017-11-01

    Valve-in-valve transcatheter aortic valve replacement is less effective in small surgical bioprostheses. We evaluated the feasibility of bioprosthetic valve fracture with a high-pressure balloon to facilitate valve-in-valve transcatheter aortic valve replacement. In vitro bench testing on aortic tissue valves was performed on 19-mm and 21-mm Mitroflow (Sorin, Milan, Italy), Magna and Magna Ease (Edwards Lifesciences, Irvine, CA), Trifecta and Biocor Epic (St. Jude Medical, Minneapolis, MN), and Hancock II and Mosaic (Medtronic, Minneapolis, MN). High-pressure balloons Tru Dilation, Atlas Gold, and Dorado (C.R. Bard, Murray Hill, NJ) were used to determine which valves could be fractured and at what pressure fracture occurred. Mitroflow, Magna, Magna Ease, Mosaic, and Biocor Epic surgical valves were successfully fractured using high-pressures balloon 1 mm larger than the labeled valve size whereas Trifecta and Hancock II surgical valves could not be fractured. Only the internal valve frame was fractured, and the sewing cuff was never disrupted. Manufacturer's rated burst pressures for balloons were exceeded, with fracture pressures ranging from 8 to 24 atmospheres depending on the surgical valve. Testing further demonstrated that fracture facilitated the expansion of previously constrained, underexpanded transcatheter valves (both balloon and self-expanding) to the manufacturer's recommended size. Bench testing demonstrates that the frame of most, but not all, bioprosthetic surgical aortic valves can be fractured using high-pressure balloons. The safety of bioprosthetic valve fracture to optimize valve-in-valve transcatheter aortic valve replacement in small surgical valves requires further clinical investigation. Copyright © 2017 The Society of Thoracic Surgeons. Published by Elsevier Inc. All rights reserved.

  15. A comparison of conventional surgery, transcatheter aortic valve replacement, and sutureless valves in "real-world" patients with aortic stenosis and intermediate- to high-risk profile.

    Science.gov (United States)

    Muneretto, Claudio; Alfieri, Ottavio; Cesana, Bruno Mario; Bisleri, Gianluigi; De Bonis, Michele; Di Bartolomeo, Roberto; Savini, Carlo; Folesani, Gianluca; Di Bacco, Lorenzo; Rambaldini, Manfredo; Maureira, Juan Pablo; Laborde, Francois; Tespili, Maurizio; Repossini, Alberto; Folliguet, Thierry

    2015-12-01

    We sought to investigate the clinical outcomes of patients with isolated severe aortic stenosis and an intermediate- to high-risk profile treated by means of conventional surgery (surgical aortic valve replacement), sutureless valve implantation, or transcatheter aortic valve replacement in a multicenter evaluation. Among 991 consecutive patients with isolated severe aortic stenosis and an intermediate- to high-risk profile (Society of Thoracic Surgeons score >4 and logistic European System for Cardiac Operative Risk Evaluation I >10), a propensity score analysis was performed on the basis of the therapeutic strategy: surgical aortic valve replacement (n = 204), sutureless valve implantation (n = 204), and transcatheter aortic valve replacement (n = 204). Primary end points were 30-day mortality and overall survival at 24-month follow-up; the secondary end point was survival free from a composite end point of major adverse cardiac events (defined as cardiac-related mortality, myocardial infarction, cerebrovascular accidents, and major hemorrhagic events) and periprosthetic regurgitation greater than 2. Thirty-day mortality was significantly higher in the transcatheter aortic valve replacement group (surgical aortic valve replacement = 3.4% vs sutureless = 5.8% vs transcatheter aortic valve replacement = 9.8%; P = .005). The incidence of postprocedural was 3.9% in asurgical aortic valve replacement vs 9.8% in sutureless vs 14.7% in transcatheter aortic valve replacement (Prisk factor for overall mortality hazard ratio (hazard ratio, 2.5; confidence interval, 1.1-4.2; P = .018). The use of transcatheter aortic valve replacement in patients with an intermediate- to high-risk profile was associated with a significantly higher incidence of perioperative complications and decreased survival at short- and mid-term when compared with conventional surgery and sutureless valve implantation. Copyright © 2015 The American Association for Thoracic Surgery. Published by

  16. Pulmonary valve stenosis

    Science.gov (United States)

    ... surgery - discharge Images Heart valves References Carabello BA. Valvular heart disease. In: Goldman L, Schafer AI, eds. Goldman's Cecil ... Saunders; 2016:chap 69. Otto CM, Bownow RO. Valvular heart disease. In: Mann DL, Zipes DP, Libby P, Bonow ...

  17. Mitral valve regurgitation

    Science.gov (United States)

    ... and dentist if you have a history of heart valve disease or congenital heart disease before treatment. Some people ... the middle Heart, front view References Carabello BA. Valvular heart disease. In: Goldman L, Schafer AI, eds. Goldman-Cecil ...

  18. Aortic Valve Disease

    Science.gov (United States)

    ... team will discuss with you the advantages and disadvantages of both valve types. Regardless of which type ... Diagnosis and Treatment Options Recovery Questions for Your Doctor Will my condition ever get better without treatment? ...

  19. Dry product valve

    International Nuclear Information System (INIS)

    Greaves, James D.

    1984-01-01

    This invention provides a system for delivering particulate radioactive or other toxic wastes to a container in which they can be solidified. The system includes a set of valves that prevent the escape of dusty materials to the atmosphere

  20. Ball check valve

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1978-01-01

    A pressurized nuclear reactor having an instrument assembly sheathed in a metallic tube which is extended vertically upward into the reactor core by traversing a metallic guide tube which is welded to the wall of the vessel is described. Sensors in each instrument assembly are connected to instruments outside the vessel to manifest the conditions within the core. Each instrument assembly probe is moved into position within a metallic guide channel. The guide channel penetrates the wall of the vessel and forms part of the barrier to the environment within the pressure vessel. Each channel includes a ball check valve which is opened by the instrument assembly probe when the probe passes through the valve. A ball valve element is moved from its seat by the probe to a position lateral of the bore of the channel and is guided to its seat along a sloped path within the valve body when the probe is removed. 5 claims, 3 figures

  1. Coanda effect in valves

    Directory of Open Access Journals (Sweden)

    Uruba Václav

    2017-01-01

    Full Text Available Coanda effect takes place in flow within valves diffuser for certain conditions. The valve plug in half-closed position forms wall-jet, which could be stable or instable, depending on geometry and other conditions. This phenomenon was subject of experimental study using time-resolved PIV technique. For the acquired data analysis the special spatio-temporal methods have been used.

  2. Bicuspid aortic valves: Diagnostic accuracy of standard axial 64-slice chest CT compared to aortic valve image plane ECG-gated cardiac CT

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, David J., E-mail: david.murphy@st-vincents.ie [Department of Radiology, St Vincent' s University Hospital, Elm Park, Dublin 4 (Ireland); McEvoy, Sinead H., E-mail: s.mcevoy@st-vincents.ie [Department of Radiology, St Vincent' s University Hospital, Elm Park, Dublin 4 (Ireland); Iyengar, Sri, E-mail: sri.iyengar@nhs.net [Department of Radiology, Plymouth Hospitals NHS Trust, Plymouth Devon PL6 8DH (United Kingdom); Feuchtner, Gudrun, E-mail: Gudrun.Feuchtner@i-med.ac.at [Department of Radiology, Innsbruck Medical University, Anichstr. 35, A-6020 Innsbruck (Austria); Cury, Ricardo C., E-mail: r.cury@baptisthealth.net [Department of Radiology, Baptist Cardiac and Vascular Institute, 8900 North Kendall Drive, Miami, FL 33176 (United States); Roobottom, Carl, E-mail: carl.roobottom@nhs.net [Department of Radiology, Plymouth Hospitals NHS Trust, Plymouth Devon PL6 8DH (United Kingdom); Plymouth University Peninsula Schools of Medicine and Dentistry (United Kingdom); Baumueller, Stephan, E-mail: Hatem.Alkadhi@usz.ch [Institute for Diagnostic and Interventional Radiology, University Hospital Zurich, Raemistrasse 100, CH-8091 Zurich (Switzerland); Alkadhi, Hatem, E-mail: stephan.baumueller@usz.ch [Institute for Diagnostic and Interventional Radiology, University Hospital Zurich, Raemistrasse 100, CH-8091 Zurich (Switzerland); Dodd, Jonathan D., E-mail: jonniedodd@gmail.com [Department of Radiology, St Vincent' s University Hospital, Elm Park, Dublin 4 (Ireland)

    2014-08-15

    Objectives: To assess the diagnostic accuracy of standard axial 64-slice chest CT compared to aortic valve image plane ECG-gated cardiac CT for bicuspid aortic valves. Materials and methods: The standard axial chest CT scans of 20 patients with known bicuspid aortic valves were blindly, randomly analyzed for (i) the appearance of the valve cusps, (ii) the largest aortic sinus area, (iii) the longest aortic cusp length, (iv) the thickest aortic valve cusp and (v) valve calcification. A second blinded reader independently analyzed the appearance of the valve cusps. Forty-two age- and sex-matched patients with known tricuspid aortic valves were used as controls. Retrospectively ECG-gated cardiac CT multiphase reconstructions of the aortic valve were used as the gold-standard. Results: Fourteen (21%) scans were scored as unevaluable (7 bicuspid, 7 tricuspid). Of the remainder, there were 13 evaluable bicuspid valves, ten of which showed an aortic valve line sign, while the remaining three showed a normal Mercedes-Benz appearance owing to fused valve cusps. The 35 evaluable tricuspid aortic valves all showed a normal Mercedes-Benz appearance (P = 0.001). Kappa analysis = 0.62 indicating good interobserver agreement for the aortic valve cusp appearance. Aortic sinus areas, aortic cusp lengths and aortic cusp thicknesses of ≥3.8 cm{sup 2}, 3.2 cm and 1.6 mm respectively on standard axial chest CT best distinguished bicuspid from tricuspid aortic valves (P < 0.0001 for all). Of evaluable scans, the sensitivity, specificity, positive and negative predictive values of standard axial chest CT in diagnosing bicuspid aortic valves was 77% (CI 0.54–1.0), 100%, 100% and 70% respectively. Conclusion: The aortic valve is evaluable in approximately 80% of standard chest 64-slice CT scans. Bicuspid aortic valves may be diagnosed on evaluable scans with good diagnostic accuracy. An aortic valve line sign, enlarged aortic sinuses and elongated, thickened valve cusps are specific CT

  3. Demonstration test for reliability of valves for atomic power plants

    International Nuclear Information System (INIS)

    Hosaka, Shiro

    1978-01-01

    The demonstration test on the reliability of valves for atomic power plants being carried out by the Nuclear Engineering Test Center is reported. This test series is conducted as six-year project from FY 1976 to FY 1981 at the Isogo Test Center. The demonstration test consists of (1) environmental test, (2) reaction force test, (3) vibration test, (4) stress measurement test, (5) operational characteristic test, (6) flow resistance coefficient measuring test, (7) leakage test and (8) safety valve and relief valve test. These contents are explained about the special requirements for nuclear use, for example, the enviornmental condition after the design base accident of PWRs and BWRs, the environmental test sequence for isolation valves of containment vessels under the emergency condition, the seismic test condition for valves of nuclear use, the various stress measurements under thermal transient conditions, the leak test after 500 cycles between the normal operating conditions for PWRs and BWRs and the start up conditions and so on. As for the testing facilities, the whole flow diagram is shown, in which the environmental test section, the vibration test section, the steam test section, the hot water test section, the safety valve test section and main components are included. The specifications of each test section and main components are presented. (Nakai, Y.)

  4. Isolation valve control device for nuclear power plant

    International Nuclear Information System (INIS)

    Yukinori, Shigeru.

    1990-01-01

    The present invention provides an isolation valve control device for detecting pipeline rupture accidents in a BWR type nuclear power plant at an early stage to close an isolation valve thereby reducing the amout of radioactivity released to the circumstance. That is, isolation valves are disposed in the pipeline for each of the systems in the nuclear power plant and flow ratemeters are disposed to at least two positions in each of the pipelines. If a meaningful difference is shown for the measured values by these flow ratemeters, the isolation valve is closed. In this way, if pipeline rupture such as leak before break (LBB) is caused to a portion of a system pipelines, the measured value from the flow ratemeters at the downstream of the pipeline is lowered. Accordingly, when a meaningful difference is formed between the value of the flow ratematers at the upstream and the downstream, occurrence of pipe rutpture between both of the flow ratemeters can be detected. As a result, the isolation valves of the system can be closed. According to the present invention, it is possible to detect the pipeline rupture at an early stage irrespective of the kind of the systems, diameter of the pipelines and the magnitude of the ruptured area, and the isolation valve can be closed. (I.S.)

  5. Fabricating microfluidic valve master molds in SU-8 photoresist

    Science.gov (United States)

    Dy, Aaron J.; Cosmanescu, Alin; Sluka, James; Glazier, James A.; Stupack, Dwayne; Amarie, Dragos

    2014-05-01

    Multilayer soft lithography has become a powerful tool in analytical chemistry, biochemistry, material and life sciences, and medical research. Complex fluidic micro-circuits require reliable components that integrate easily into microchips. We introduce two novel approaches to master mold fabrication for constructing in-line micro-valves using SU-8. Our fabrication techniques enable robust and versatile integration of many lab-on-a-chip functions including filters, mixers, pumps, stream focusing and cell-culture chambers, with in-line valves. SU-8 created more robust valve master molds than the conventional positive photoresists used in multilayer soft lithography, but maintained the advantages of biocompatibility and rapid prototyping. As an example, we used valve master molds made of SU-8 to fabricate PDMS chips capable of precisely controlling beads or cells in solution.

  6. Fabricating microfluidic valve master molds in SU-8 photoresist

    International Nuclear Information System (INIS)

    Dy, Aaron J; Cosmanescu, Alin; Sluka, James; Glazier, James A; Amarie, Dragos; Stupack, Dwayne

    2014-01-01

    Multilayer soft lithography has become a powerful tool in analytical chemistry, biochemistry, material and life sciences, and medical research. Complex fluidic micro-circuits require reliable components that integrate easily into microchips. We introduce two novel approaches to master mold fabrication for constructing in-line micro-valves using SU-8. Our fabrication techniques enable robust and versatile integration of many lab-on-a-chip functions including filters, mixers, pumps, stream focusing and cell-culture chambers, with in-line valves. SU-8 created more robust valve master molds than the conventional positive photoresists used in multilayer soft lithography, but maintained the advantages of biocompatibility and rapid prototyping. As an example, we used valve master molds made of SU-8 to fabricate PDMS chips capable of precisely controlling beads or cells in solution. (technical note)

  7. Tchernobyl accident

    International Nuclear Information System (INIS)

    1986-06-01

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

  8. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  9. Fixation and mounting of porcine aortic valves for use in mock circuits.

    Science.gov (United States)

    Schlöglhofer, Thomas; Aigner, Philipp; Stoiber, Martin; Schima, Heinrich

    2013-10-01

    Investigations of the circulatory system in vitro use mock circuits that require valves to mimic the cardiac situation. Whereas mechanical valves increase water hammer effects due to inherent stiffness and do not allow the use of pressure lines or catheters, bioprosthetic valves are expensive and of limited durability in test fluids. Therefore, we developed a cheap, fast, alternative method to mount valves obtained from the slaughterhouse in mock circuits. Porcine aortic roots were obtained from the abattoir and used either in native condition or after fixation. Fixation was performed at a constant retrograde pressure to ensure closed valve position. Fixation time was 4 h in a 0.5%-glutaraldehyde phosphate buffer. The fixed valves were molded into a modular mock circulation connector using a fast curing silicone. Valve functionality was evaluated in a pulsatile setting (cardiac output = 4.7 l/min, heart rate = 80 beats/min) and compared before and after fixation. Leaflet motion was recorded with a high-speed camera and valve insufficiency was quantified by leakage flow under steady pressure application (80 mmHg). Under physiological conditions the aortic valves showed almost equal leaflet motion in native and fixed conditions. However, the leaflets of the native valves showed lower stiffness and more fluttering during systole than the fixed specimens. Under retrograde pressure, fresh and fixed valves showed small leakage flows of <30 ml/min. The new mounting and fixation procedure is a fast method to fabricate low cost biologic valves for the use in mock circuits.

  10. Transcatheter Aortic Valve Replacement for Degenerative Bioprosthetic Surgical Valves

    DEFF Research Database (Denmark)

    Dvir, Danny; Webb, John; Brecker, Stephen

    2012-01-01

    Transcatheter aortic valve-in-valve implantation is an emerging therapeutic alternative for patients with a failed surgical bioprosthesis and may obviate the need for reoperation. We evaluated the clinical results of this technique using a large, worldwide registry....

  11. NRC valve performance test program - check valve testing

    International Nuclear Information System (INIS)

    Jeanmougin, N.M.

    1987-01-01

    The Valve Performance Test Program addresses the current requirements for testing of pressure isolation valves (PIVs) in light water reactors. Leak rate monitoring is the current method used by operating commercial power plants to survey the condition of their PIVs. ETEC testing of three check valves (4-inch, 6-inch, and 12-inch nominal diameters) indicates that leak rate testing is not a reliable method for detecting impending valve failure. Acoustic emission monitoring of check valves shows promise as a method of detecting loosened internals damage. Future efforts will focus on evaluation of acoustic emission monitoring as a technique for determining check valve condition. Three gate valves also will be tested to evaluate whether the check valve results are applicable to gate type PIVs

  12. Prosthetic valve obstruction: Redo surgery or fibrinolysis?

    Directory of Open Access Journals (Sweden)

    Avinash Inamdar

    2017-01-01

    Full Text Available Objective: The aim of this study was to compare the efficacy and safety of surgery versus fibrinolytic therapy in patients with prosthetic valve obstruction. Materials and Methods: We compared 15 patients of prosthetic valve thrombosis treated by surgical line of management and another 15 patients treated by thrombolysis. All patients were initially assessed by clinical evaluation and diagnosis confirmed by transthoracic and transesophageal two-dimensional echocardiography. Depending on hemodynamic stability, pannus, or thrombus on transesophageal echocardiography, the patients were assigned surgical or medical line of management. Results: Patients mortality rate was 40% in fibrinolytic group and 13.33% in surgical group. Recurrence was 40% in fibrinolytic group while there was no recurrence till date in surgery group. Complications were more in fibrinolytic group as opposed to surgery group patient. Conclusion: From our experience, we conclude that redo surgery is effective and definitive treatment, especially in patients with stable hemodynamic conditions.

  13. Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

    International Nuclear Information System (INIS)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I.; Elkin, I.V.

    2001-01-01

    The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG and begins to remove

  14. Investigation of primary-to-secondary leakage accident on the PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Lipatov, I.A.; Dremin, G.I.; Galtchanskaya, S.A.; Chmal, I.I.; Moloshnikov, A.S.; Gorbunov, Y.S.; Antonova, A.I. [Electrogorsk Research and Engineering Center, EREC, Moscow (Russian Federation); Elkin, I.V. [RRC ' ' Kurchatov Institute, Moscow (Russian Federation)

    2001-07-01

    The full text follows. The paper presents the main results from the test on primary-to-secondary leakage of 100 mm in equivalent diameter. The test was performed on the PSB-VVER integral test facility. PSB-VVER is a 4-loops scaled down model of primary system of NPP with VVER-1000 Russian type reactor. Volume - power scale is about 1/300 while elevation scale is 1/1. All components of the primary system of the reference NPP are modeled on PSB-VVER. Both passive (accumulators) and active (high and low pressure) ECCSs, pressurizer spray and relief circuits, feed water system and atmospheric dumping system (ADS) as well as the primary circuit gas remove emergency system are also simulated. The primary-to-secondary leakage was simulated using an external break line which connects the upper part of the hot header to SG water volume. The break line included a break nozzle (a cylindrical channel d = 5.8 mm, l/d = 10 with sharp inlet edge), quick-acting valve and two-phase mass flow rate measurement system. In addition loss of off-site power at the moment when a scram-signal is generated was assumed in the experiment. Thus the accident is to be considered as a beyond-design-basic one. The loss of off-site power results in the following: -main circulation pump shutdown; -pressurizer heaters switching off; -HPIS water cooling flow rate and number of points of water injection are reduced The study focuses on the adequacy of the associated accident management (AM) procedure developed by EDO ''GIDROPRESS'' as a General Designer of VVER-type reactors. The AM-procedure was adopted to the PSB-VVER test facility conditions using CATHARE (France) and DINAMIKA (Russia) codes analysis. The AM-procedure in PSB-VVER is as follows: after about 30 min of the onset of the accident, when the accident type and the localization of the SG affected become evident for the operator, he closes all the main steam isolation valves, inhibits the ADS actuation in the affected SG

  15. Check valves aging assessment

    International Nuclear Information System (INIS)

    Haynes, H.D.

    1991-01-01

    In support of the NRC Nuclear Plant Aging Research (NPAR) program, the Oak Ridge National Laboratory (ORNL) has carried out an assessment of several check value diagnostic monitoring methods, in particular, those based on measurements of acoustic emission, ultrasonics, and magnetic flux. The evaluations have focussed on the capabilities of each method to provide information useful in determining check valve aging and service wear effects, check valve failures, and undesirable operating modes. This paper describes the benefits and limitations associated with each method and includes recent laboratory and field test data, including data obtained from the vendors who recently participated in a comprehensive series of tests directed by a nuclear industry users group. In addition, as part of the ORNL Advanced Diagnostic Engineering Research and Development Center (ADEC), two novel nonintrusive monitoring methods were developed that provide several unique capabilities. These methods, based on external ac- an dc-magnetic monitoring are also described. None of the examined methods could, by themselves, monitor both the instantaneous position and motion of check valve internals and valve leakage; however, the combination of acoustic emission monitoring with one of the other methods provides the means to determine vital check valve operational information

  16. On the testing fast response NPP's valves of large nominal bores and high parameters

    International Nuclear Information System (INIS)

    Majorov, A.P.; Ostretsov, I.N.

    1990-01-01

    Investigation technique for valves of large norminal bores and high parameters which is based on application of simulation effect for operation and accident loadings during movement of valve lock at bench tests with medium flow rate by 100-1000 times less than during operation is given. Loading simulation technique is provided using simulator of lock loading. Investigation results are essential to make decision concerning advisability of serial production of fittings without full-scale test conducting

  17. Risk evaluation for motor operated valves in an Inservice Testing Program at a PWR nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    Li, Y.C.; Chen, K.T.; Su, Y.L.; Ting, K.; Chien, F.T.; Li, G.D.; Huang, S.H.

    2012-01-01

    Safety related valves such as Motor Operated Valves (MOV), Air Operated Valves (AOV) or Check Valves (CV) play an important role in nuclear power plant. Functioning of these valves mainly aim at emergency reactivity control, post-accident residue heat removal, post-accident radioactivity removal and containment isolation when a design basis accident occurred. In order to maintain these valves under operable conditions, an Inservice Testing Program (IST) is defined for routine testing tasks based on the ASME Boiler and Pressure Vessel Code section XI code requirements. Risk based Inservice Testing Programs have been studied and developed extensively in the nuclear energy industry since the 1990s. Risk Based evaluations of IST can bring positive advantages to the licensee such as identifying the vulnerability of the system, reducing unnecessary testing burden, concentrating testing resources on the critical pass oriented valves and saving plant’s personnel dose exposure. This risk evaluation is incorporated with quantitative and qualitative analyses to the Motor Operated Valves under current Inservice Testing Program for PWR nuclear power plant in Taiwan. With the outcome of the risk classifications for the safety related MOVs through numerical or deterministic analyses, a risk based testing frequency relief is suggested to demonstrate the benefits received from the risk based Inservice Testing Program. The goal made of this study, it could be as a reference and cornerstone for the licensee to perform overall scope Risk-Informed Inservice Testing Program (RI-IST) evaluation by referring relevant methodologies established in this study.

  18. Prevention of pedestrian accidents.

    OpenAIRE

    Kendrick, D

    1993-01-01

    Child pedestrian accidents are the most common road traffic accident resulting in injury. Much of the existing work on road traffic accidents is based on analysing clusters of accidents despite evidence that child pedestrian accidents tend to be more dispersed than this. This paper analyses pedestrian accidents in 573 children aged 0-11 years by a locally derived deprivation score for the years 1988-90. The analysis shows a significantly higher accident rate in deprived areas and a dose respo...

  19. Development of a discharge model for the Bopp and Reuther Degasser/Condenser relief valves for heat sink assessment

    International Nuclear Information System (INIS)

    Hasnaoui, C. . chiheb@hasnaoui.net; Huynh, M.

    2004-01-01

    A total loss of all sustained engineering heat sinks is considered as a severe accident with low probability of occurrence. Following a total loss of all sustained engineering heat sinks, the Degasser/Condenser relief valves (3332-RV11 and RV21) would then become the sole means available for the depressurization of the primary heat transport system. Accurate estimation of the discharge through these valves is required to assess the impact of this kind of accident on fuel cooling and the primary circuit integrity. This paper describes a model used to estimate the Degasser/Condenser relief valve discharge capacity. This model is used to predict the flow discharge under a range of conditions upstream of the relief valves; from sub-cooled to saturated liquid and up to vapor conditions. The defined model is then used to estimate the relief valve discharge rates under various hypothetical conditions of the PHTS using the Cathena code. (author)

  20. Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

    International Nuclear Information System (INIS)

    Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

    1995-01-01

    Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

  1. Aortic valve replacement and the stentless Freedom SOLO valve

    NARCIS (Netherlands)

    Wollersheim, L.W.L.M.

    2016-01-01

    Aortic valve stenosis has become the most prevalent valvular heart disease in Europe and North America, and is generally caused by age-related calcification of the aortic valve. For most patients, severe symptomatic aortic stenosis needs effective mechanical relief in the form of valve replacement

  2. Comparative study of Butterfly valves

    International Nuclear Information System (INIS)

    Galmes Belmonte, F.B.

    1998-01-01

    This work tries to justify the hydrodynamic butterfly valves performance, using the EPRI tests, results carried out in laboratory and in situ. This justification will be possible if: - The valves to study are similar - Their performance is calculated using EPRI's methodology Looking for this objective, the elements of the present work are: 1. Brief EPRI butterfly valve description it wild provide the factors which are necessary to define the butterfly valves similarity. 2. EPRI tests description and range of validation against test data definition. 3. Description of the spanish butterfly analyzed valves, and comparison with the EPRI performance results, to prove that this valves are similar to the EPRI test valves. In this way, it will not be necessary to carry out particular dynamic tests on the spanish valves to describe their hydrodynamic performance. (Author)

  3. A symmetric safety valve

    International Nuclear Information System (INIS)

    Burtraw, Dallas; Palmer, Karen; Kahn, Danny

    2010-01-01

    How to set policy in the presence of uncertainty has been central in debates over climate policy. Concern about costs has motivated the proposal for a cap-and-trade program for carbon dioxide, with a 'safety valve' that would mitigate against spikes in the cost of emission reductions by introducing additional emission allowances into the market when marginal costs rise above the specified allowance price level. We find two significant problems, both stemming from the asymmetry of an instrument that mitigates only against a price increase. One is that most important examples of price volatility in cap-and-trade programs have occurred not when prices spiked, but instead when allowance prices collapsed. Second, a single-sided safety valve may have unintended consequences for investment. We illustrate that a symmetric safety valve provides environmental and welfare improvements relative to the conventional one-sided approach.

  4. Nuclear reactor steam depressurization valve

    International Nuclear Information System (INIS)

    Moore, G.L.

    1991-01-01

    This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

  5. Valve spindle gland

    International Nuclear Information System (INIS)

    Burda, Z.; Harazim, A.; Kerlin, K.

    1979-01-01

    A gland is proposed of the valve spindle designed for radioactive or otherwise harmful media, such as in nuclear power plant primary circuits. The gland is installed in the valve cover and consists of a primary and a secondary part and of a gland case partitioning the gland space into two chambers. The bottom face of the gland case is provided with a double-sided collar for controlling the elements of the bottom primary gland while the top face is provided with a removable flange. (M.S.)

  6. Building valve amplifiers

    CERN Document Server

    Jones, Morgan

    2013-01-01

    Building Valve Amplifiers is a unique hands-on guide for anyone working with tube audio equipment--as an electronics hobbyist, audiophile or audio engineer. This 2nd Edition builds on the success of the first with technology and technique revisions throughout and, significantly, a major new self-build project, worked through step-by-step, which puts into practice the principles and techniques introduced throughout the book. Particular attention has been paid to answering questions commonly asked by newcomers to the world of the valve, whether audio enthusiasts tackling their first build or

  7. Valve thrombosis following transcatheter aortic valve implantation: a systematic review.

    Science.gov (United States)

    Córdoba-Soriano, Juan G; Puri, Rishi; Amat-Santos, Ignacio; Ribeiro, Henrique B; Abdul-Jawad Altisent, Omar; del Trigo, María; Paradis, Jean-Michel; Dumont, Eric; Urena, Marina; Rodés-Cabau, Josep

    2015-03-01

    Despite the rapid global uptake of transcatheter aortic valve implantation, valve trombosis has yet to be systematically evaluated in this field. The aim of this study was to determine the clinical characteristics, diagnostic criteria, and treatment outcomes of patients diagnosed with valve thrombosis following transcatheter aortic valve implantation through a systematic review of published data. Literature published between 2002 and 2012 on valve thrombosis as a complication of transcatheter aortic valve implantation was identified through a systematic electronic search. A total of 11 publications were identified, describing 16 patients (mean age, 80 [5] years, 65% men). All but 1 patient (94%) received a balloon-expandable valve. All patients received dual antiplatelet therapy immediately following the procedure and continued to take either mono- or dual antiplatelet therapy at the time of valve thrombosis diagnosis. Valve thrombosis was diagnosed at a median of 6 months post-procedure, with progressive dyspnea being the most common symptom. A significant increase in transvalvular gradient (from 10 [4] to 40 [12] mmHg) was the most common echocardiographic feature, in addition to leaflet thickening. Thrombus was not directly visualized with echocardiography. Three patients underwent valve explantation, and the remaining received warfarin, which effectively restored the mean transvalvular gradient to baseline within 2 months. Systemic embolism was not a feature of valve thrombosis post-transcatheter aortic valve implantation. Although a rare, yet likely under-reported complication of post-transcatheter aortic valve implantation, progressive dyspnea coupled with an increasing transvalvular gradient on echocardiography within the months following the intervention likely signifies valve thrombosis. While direct thrombus visualization appears difficult, prompt initiation of oral anticoagulation therapy effectively restores baseline valve function. Copyright © 2014

  8. Use of Main Loop Isolating Valves (GZZS) in WWER 440

    International Nuclear Information System (INIS)

    Stefanova, A.E.; Gencheva, R.V.; Groudev, P.P.

    2002-01-01

    This paper discusses the usage of Main Loop Isolation Valves in case of Steam Generator Tube Rupture accident in WWER440/V230. A double-ended single pipe break in SG-6 was chosen as representative. In the paper are investigated two cases. In the first one the operator isolates the affected loop by Main Loop Isolation Valves closing and after primary depressurization re-opens them to cooldown the damaged Steam Generator. The second case treats the situation, where Main Loop Isolation Valves fail to close with the necessary operator actions for managing plant recovery. RELAP5/MOD3.2 computer code has been used to simulate the Steam Generator Tube Rupture accident in WWER440 NPP model. This model was developed and validated at Institute for Nuclear Research and Nuclear Energy - Bulgarian Academy of Sciences. The results of analyses presented in this report demonstrate that in the both cases (with or without Main Loop Isolation Valves usage) the operator could bring the plant to stable and safety conditions (Authors)

  9. Cavitation problems in sodium valves

    International Nuclear Information System (INIS)

    Elie, X.

    1976-01-01

    Cavitation poses few problems for sodium valves, in spite of the fact that the loops are not pressurized. This is no doubt due to the low flow velocities in the pipes. For auxiliary loop valves we are attempting to standardize performances with respect to cavitation. For economic reasons cavitation thresholds are approached with large diameter valves. (author)

  10. Valve Concepts for Microfluidic Cell Handling

    Directory of Open Access Journals (Sweden)

    M. Grabowski

    2010-01-01

    Full Text Available In this paper we present various pneumatically actuated microfluidic valves to enable user-defined fluid management within a microfluidic chip. To identify a feasible valve design, certain valve concepts are simulated in ANSYS to investigate the pressure dependent opening and closing characteristics of each design. The results are verified in a series of tests. Both the microfluidic layer and the pneumatic layer are realized by means of soft-lithographic techniques. In this way, a network of channels is fabricated in photoresist as a molding master. By casting these masters with PDMS (polydimethylsiloxane we get polymeric replicas containing the channel network. After a plasma-enhanced bonding process, the two layers are irreversibly bonded to each other. The bonding is tight for pressures up to 2 bar. The valves are integrated into a microfluidic cell handling system that is designed to manipulate cells in the presence of a liquid reagent (e.g. PEG – polyethylene glycol, for cell fusion. For this purpose a user-defined fluid management system is developed. The first test series with human cell lines show that the microfluidic chip is suitable for accumulating cells within a reaction chamber, where they can be flushed by a liquid medium.

  11. SWR-1000 concept on control of severe accidents

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1998-01-01

    It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

  12. Radiation accidents

    International Nuclear Information System (INIS)

    Saenger, E.L.

    1986-01-01

    It is essential that emergency physicians understand ways to manage patients contaminated by radioactive materials and/or exposed to external radiation sources. Contamination accidents require careful surveys to identify the metabolic pathway of the radionuclides to guide prognosis and treatment. The level of treatment required will depend on careful surveys and meticulous decontamination. There is no specific therapy for the acute radiation syndrome. Prophylactic antibodies are desirable. For severely exposed patients treatment is similar to the supportive care given to patients undergoing organ transplantation. For high-dose extremity injury, no methods have been developed to reverse the fibrosing endarteritis that eventually leads to tissue death so frequently found with this type of injury. Although the Three Mile Island episode of March 1979 created tremendous public concern, there were no radiation injuries. The contamination outside the reactor building and the release of radioiodine were negligible. The accidental fuel element meltdown at Chernobyl, USSR, resulted in many cases of acute radiation syndrome. More than 100,000 people were exposed to high levels of radioactive fallout. The general principles outlined here are applicable to accidents of that degree of severity

  13. Chernobyl accident

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1995-01-01

    The monograph contains the catastrophe's events chronology, the efficiency assessed of those measures assumed for their localization as well as their environmental and socio-economic impact. Among materials of the monograph the results are presented of research on the radioactive contamination field forming as well as those concerning the investigation of biogeochemical properties of Chernobyl radionuclides and their migration process in the environment of the Ukraine. The data dealing with biological effects of the continued combined internal and external radioactive influence on plants, animals and human health under the circumstances of Chernobyl accident are of the special interest. In order to provide the scientific generalizing information on the medical aspects of Chernobyl catastrophe, the great part of the monograph is allotted to appraise those factors affecting the health of different population groups as well as to depict clinic aspects of Chernobyl events and medico-sanitarian help system. The National Programme of Ukraine for the accident consequences elimination and population social protection assuring for the years 1986-1993 and this Programme concept for the period up to the year 2000 with a special regard of the world community participation there

  14. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

  15. Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

  16. Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

    International Nuclear Information System (INIS)

    Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

    2016-01-01

    The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

  17. The Fukushima accident

    International Nuclear Information System (INIS)

    Maqua, M.; Stueck, R.

    2012-01-01

    On 11 March 2011, the Tohoku earthquake and the subsequent tsunami hit the Japanese east coast, causing more than 15,000 fatalities. To this date, 3,000 people are still missing. The Fukushima Dai-ichi NPP was the nuclear installation that was most affected by the tsunami. The earthquake cut off the NPP from the national grid. About 45 minutes later, the tsunami flooded units 1-4 and led to core meltdown events with large releases for units 1, 2 and 3. Unit 4 had been in refuelling outage at that time and lost the cooling of the spent fuel pool for several days. Considerable hydrogen explosions occurred in units 1, 3 and 4. Shortly after the accident, TEPCO started to mitigate the consequences of the accident by providing external cooling to the reactors and by removing the radioactive debris from the site. Great emphasis was laid on effective radiation protection measures for the clean-up workers. Thus, up to now there has been no fatality due to the radiation caused by the Fukushima accident. The main steps of the accident sequences are described, taking into account the latest findings of investigations performed by TEPCO or on behalf of the regulatory body. The presentation focuses on the description of the status of the Fukushima Dai-ichi nuclear power plant and the future steps for cleaning-up the site. In the presentation, the major phases of the roadmap that TEPCO has developed for the clean-up are highlighted. The risks associated with the current plant status and the clean-up phases are described. Abstract the content of the manuscript in a few lines.

  18. Transcatheter aortic valve implantation in failed bioprosthetic surgical valves

    DEFF Research Database (Denmark)

    Dvir, Danny; Webb, John G; Bleiziffer, Sabine

    2014-01-01

    for patients with structural valve deterioration; however, a comprehensive evaluation of survival after the procedure has not yet been performed. OBJECTIVE: To determine the survival of patients after transcatheter valve-in-valve implantation inside failed surgical bioprosthetic valves. DESIGN, SETTING......, stroke, and New York Heart Association functional class. RESULTS: Modes of bioprosthesis failure were stenosis (n = 181 [39.4%]), regurgitation (n = 139 [30.3%]), and combined (n = 139 [30.3%]). The stenosis group had a higher percentage of small valves (37% vs 20.9% and 26.6% in the regurgitation...... and combined groups, respectively; P = .005). Within 1 month following valve-in-valve implantation, 35 (7.6%) patients died, 8 (1.7%) had major stroke, and 313 (92.6%) of surviving patients had good functional status (New York Heart Association class I/II). The overall 1-year Kaplan-Meier survival rate was 83...

  19. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1981-01-01

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  20. Development of a butterfly check valve model under natural circulation conditions

    International Nuclear Information System (INIS)

    Rao, Yuxian; Yu, Lei; Fu, Shengwei; Zhang, Fan

    2015-01-01

    Highlights: • Bases on Lim’s swing check valve model, a butterfly check valve model was developed. • The method to quantify the friction torque T F in Li’s model was corrected. • The developed model was implemented into the RELAP5 code and verified. - Abstract: A butterfly check valve is widely used to prevent a reverse flow in the pipe lines of a marine nuclear power plant. Under some conditions, the natural circulation conditions in particular, the fluid velocity through the butterfly check valve might become too low to hold the valve disk fully open, thereby the flow resistance of the butterfly check valve varies with the location of the valve disk and as a result the fluid flow is significantly affected by the dynamic motion of the valve disk. Simulation of a pipe line that includes some butterfly check valves, especially under natural circulation conditions, is thus complicated. This paper focuses on the development of a butterfly check valve model to enhance the capability of the thermal–hydraulic system code and the developed model is implemented into the RELAP5 code. Both steady-state calculations and transient calculations were carried out for the primary loop system of a marine nuclear power plant and the calculation results are compared with the experimental data for verification purpose. The simulation results show an agreement with the experimental data

  1. Valves for condenser-cooling-water circulating piping in thermal power station and nuclear power station

    International Nuclear Information System (INIS)

    Kondo, Sumio

    1977-01-01

    Sea water is mostly used as condenser cooling water in thermal and nuclear power stations in Japan. The quantity of cooling water is 6 to 7 t/sec per 100,000 kW output in nuclear power stations, and 3 to 4 t/sec in thermal power stations. The pipe diameter is 900 to 2,700 mm for the power output of 75,000 to 1,100,000 kW. The valves used are mostly butterfly valves, and the reliability, economy and maintainability must be examined sufficiently because of their important role. The construction, number and arrangement of the valves around a condenser are different according to the types of a turbine and the condenser and reverse flow washing method. Three types are illustrated. The valves for sea water are subjected to the electrochemical corrosion due to sea water, the local corrosion due to stagnant water, the fouling by marine organisms, the cavitation due to valve operation, and the erosion by earth and sand. The fundamental construction, use and features of butterfly valves are described. The cases of the failure and repair of the valves after their delivery are shown, and they are the corrosion of valve bodies and valve seats, and the separation of coating and lining. The newly developed butterfly valve with overall water-tight rubber lining is introduced. (Kako, I.)

  2. Application of artificial intelligence to motor operated valve testing

    International Nuclear Information System (INIS)

    Bogard, T.; Bednar, F.; Matty, T.; Kent, R.

    1989-01-01

    Improper valve maintenance can be a significant roadblock to successful power plant operation. There have been events during which motor operated valves failed on demand due to improper switch settings. For nuclear electric generating stations, these events have led to regulatory requirements such as NRC Bulletin 85-03 and NRC Bulletin 89-10 Safety Related Motor Operated Valve Testing and Surveillance which imposes strict testing and programmatic requirements on motor operated valves (MOV). Part of the requirements include performing diagnostic testing to verify stem thrust loads and switch settings. Diagnostic equipment must be non-intrusive, minimize valve disassembly, and reduce plant refueling critical path time for testing. In this paper an on-line diagnostic system using sensors to measure stem forces, motor current, and valve position, and a portable system employing these same sensor inputs in addition to torque, limit and torque bypass switch inputs is described. Sophisticated graphic software is employed to display data or trace information. A rule based artificial intelligence (AI) system is used to analyze sensor outputs. Objectives for valve diagnostics, sample AI rules, results of actual field testing, and system software/hardware architecture are presented

  3. SAFETY SHUTOFF VALVE

    DEFF Research Database (Denmark)

    2010-01-01

    It is disclosed a shut-off valve which acts automatically and has a fully mechanical performance with respect to the loosing of the tower-shape part balance under the effect of the special acceleration Which is arisen from the quakes waves or serious vibrations, while such vibrations are mainly r...

  4. Heart valve surgery - discharge

    Science.gov (United States)

    ... ACC guideline for the management of patients with valvular heart disease: executive summary: a report of the American College ... Editorial team. Related MedlinePlus Health Topics Heart Surgery Heart Valve Diseases Browse the Encyclopedia A.D.A.M., Inc. ...

  5. Poppet valve tester

    Science.gov (United States)

    Tellier, G. F.

    1973-01-01

    Tester investigates fundamental factors affecting cyclic life and sealing performance of valve seats and poppets. Tester provides for varying impact loading of poppet against seat and rate of cycling, and controls amount and type of relative motion between sealing faces of seat and poppet. Relative motion between seat and poppet can be varied in three modes.

  6. Thermostatic Radiator Valve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Dentz, Jordan [Advanced Residential Integrated Energy Solutions Collaborative, New York, NY (United States); Ansanelli, Eric [Advanced Residential Integrated Energy Solutions Collaborative, New York, NY (United States)

    2015-01-01

    A large stock of multifamily buildings in the Northeast and Midwest are heated by steam distribution systems. Losses from these systems are typically high and a significant number of apartments are overheated much of the time. Thermostatically controlled radiator valves (TRVs) are one potential strategy to combat this problem, but have not been widely accepted by the residential retrofit market.

  7. Blocked Urethral Valves

    Science.gov (United States)

    ... if any damage has occurred to the upper urinary tract. Your pediatrician will consult with a pediatric nephrologist (kidney specialist) or nurologist, who may recommend surgery to remove the obstructing valves and prevent further infection or damage to the kidneys or urinary system. ...

  8. Tricuspid valve endocarditis

    Science.gov (United States)

    Hussain, Syed T.; Witten, James; Shrestha, Nabin K.; Blackstone, Eugene H.

    2017-01-01

    Right-sided infective endocarditis (RSIE) is less common than left-sided infective endocarditis (IE), encompassing only 5–10% of cases of IE. Ninety percent of RSIE involves the tricuspid valve (TV). Given the relatively small numbers of TVIE cases operated on at most institutions, the purpose of this review is to highlight and discuss the current understanding of IE involving the TV. RSIE and TVIE are strongly associated with intravenous drug use (IVDU), although pacemaker leads, defibrillator leads and vascular access for dialysis are also major risk factors. Staphylococcus aureus is the predominant causative organism in TVIE. Most patients with TVIE are successfully treated with antibiotics, however, 5–16% of RSIE cases eventually require surgical intervention. Indications and timing for surgery are less clear than for left-sided IE; surgery is primarily considered for failed medical therapy, large vegetations and septic pulmonary embolism, and less often for TV regurgitation and heart failure. Most patients with an infected prosthetic TV will require surgery. Concomitant left-sided IE has its own surgical indications. Earlier surgical intervention may potentially prevent further destruction of leaflet tissue and increase the likelihood of TV repair. Fortunately, TV debridement and repair can be accomplished in most cases, even those with extensive valve destruction, using a variety of techniques. Valve repair is advocated over replacement, particularly in IVDUs patients who are young, non-compliant and have a higher risk of recurrent infection and reoperation with valve replacement. Excising the valve without replacing, it is not advocated; it has been reported previously, but these patients are likely to be symptomatic, particularly in cases with septic pulmonary embolism and increased pulmonary vascular resistance. Patients with concomitant left-sided involvement have worse prognosis than those with RSIE alone, due predominantly to greater likelihood of

  9. Chernobyl accident

    International Nuclear Information System (INIS)

    Capra, D.; Facchini, U.; Gianelle, V.; Ravasini, G.; Bacci, P.

    1988-01-01

    The radioactive cloud released during the Chernobyl accident reached the Padana plain and Lombardy in the night of April 30th 1986; the cloud remained in the northern Italian skies for a few days and then disappeared either dispersed by winds and washed by rains. The evidence in atmosphere of radionuclides as Tellurium, Iodine, Cesium, was promptly observed. The intense rain, in first week of may, washed the radioactivity and fall-out contamined the land, soil, grass. The present work concerns the overall contamination of the Northern Italy territory and in particular the radioactive fall-out in the Lakes region. Samples of soil have been measured at the gamma spectroscope; a correlation is found between the radionuclides concentration in soil samples and the rain intensity, when appropriate deposition models are considered. A number of measurements has been done on the Como'lake ecosystem: sediments, plankton, fishes and the overall fall-out in the area has been investigated

  10. Self-reported accidents

    DEFF Research Database (Denmark)

    Møller, Katrine Meltofte; Andersen, Camilla Sloth

    2016-01-01

    The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

  11. Accident Assessment

    International Nuclear Information System (INIS)

    Tripputi, Ivo; Lund, Ingemar

    2002-01-01

    There is a general feeling that decommissioning is an activity involving limited risks, compared to NPP operation, and in particular risks involving the general public. This is technically confirmed by licensing analysis and evaluations, where, once the spent fuel has been removed from the plant, the radioactivity inventory available to be released to the environment is very limited. Decommissioning activities performed so far in the world have also confirmed the first assumptions and no specific issue has been identified, in this field, to justify a completely new approach. Commercial interests in international harmonization, which could drive an in-depth discussion about the bases of this approach, are weak at the moment. However, there are several reasons why a discussion in an international framework about the Safety Case for decommissioning (and, in particular, about Accident Assessment) may be considered necessary and important, and why it may show some specific and peculiar aspects. An effort for a comprehensive and systematic D and D accident safety assessment of the decommissioning process is justified. It is necessary also to explore in a holistic way the aspects of industrial safety, and develop tools for the decision-making process optimization. The expected results are the implementation of appropriate and optimized protective measures in any event and of adequate on/off-site emergency plans for optimal public and workers protection. The experience from other decommissioning projects and large-scale industrial activities is essential to balance provisions and an Operating Experience review process (specific for decommissioning) should help to focus on real issues

  12. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-15

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis.

  13. Simulation of A Main Steam Line Break Accident Using the Coupled 'System Thermal-Hydraulics, 3D reactor Kinetics, and Hot Channel' Analysis Capability of MARS 3.0

    International Nuclear Information System (INIS)

    Jeong, Jae Jun; Chung, Bub Dong

    2005-09-01

    For realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients. In this work, a main steam line break (MSLB) accident is simulated using the coupled 'system T-H, 3-D reactor kinetics, and hot channel analysis' feature of the MARS code. Two coupled calculations are performed for demonstration. First, a coupled calculation of the 'system T-H and 3-D reactor kinetics' with a refined core T-H nodalization is carried out to obtain global core power and local departure from nucleate boiling (DNB) ratio (DNBR) behaviors. Next, for a more accurate DNBR prediction, another coupled calculation with subchannel meshes for the hot channels is performed. The results of the coupled calculations are very reasonable and consistent so that these can be used to remove the excessive conservatism in the conventional safety analysis

  14. Diesel Engine Valve Clearance Detection Using Acoustic Emission

    Directory of Open Access Journals (Sweden)

    Fathi Elamin

    2010-01-01

    Full Text Available This paper investigated, using experimental method, the suitability of acoustic emission (AE technique for the condition monitoring of diesel engine valve faults. The clearance fault was adjusted experimentally in an exhaust valve and successfully detected and diagnosed in a Ford FSD 425 four-cylinder, four-stroke, in-line OHV, direct injection diesel engine. The effect of faulty exhaust valve clearance on engine performance was monitored and the difference between the healthy and faulty engine was observed from the recorded AE signals. The measured results from this technique show that using only time domain and frequency domain analysis of acoustic emission signals can give a superior measure of engine condition. This concludes that acoustic emission is a powerful and reliable method of detection and diagnosis of the faults in diesel engines and this is considered to be a unique approach to condition monitoring of valve performance.

  15. Large butterfly valve design copes with out-of-round pipe

    International Nuclear Information System (INIS)

    Saar, R.P.

    1975-01-01

    Two 96 inch circulating water lines at the Trojan reactor were joined to butterfly valves which had to be distorted to conform to the badly out-of-round pipes. Bubble tight seating was achieved by positioning a flexible seat ring after the valve was installed

  16. Cyclonic valve test: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Andre Sampaio; Moraes, Carlos Alberto C.; Marins, Luiz Philipe M.; Soares, Fabricio; Oliveira, Dennis; Lima, Fabio Soares de; Airao, Vinicius [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil); Ton, Tijmen [Twister BV, Rijswijk (Netherlands)

    2012-07-01

    For many years, the petroleum industry has been developing a valve that input less shear to the flow for a given required pressure drop and this can be done using the cyclonic concept. This paper presents a comparison between the performances of a cyclonic valve (low shear) and a conventional globe valve. The aim of this work is to show the advantages of using a cyclonic low shear valve instead of the commonly used in the primary separation process by PETROBRAS. Tests were performed at PETROBRAS Experimental Center (NUEX) in Aracaju/SE varying some parameters: water cut; pressure loss (from 4 kgf/cm2 to 10 kgf/cm2); flow rates (30 m3/h and 45 m3/h). Results indicates a better performance of the cyclonic valve, if compared with a conventional one, and also that the difference of the performance, is a function of several parameters (emulsion stability, water content free, and oil properties). The cyclonic valve tested can be applied as a choke valve, as a valve between separation stages (for pressure drop), or for controlling the level of vessels. We must emphasize the importance to avoid the high shear imposed by conventional valves, because once the emulsion is created, it becomes more difficult to break it. New tests are being planned to occur in 2012, but PETROBRAS is also analyzing real cases where the applications could increase the primary process efficiency. In the same way, the future installations are also being designed considering the cyclonic valve usage. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  18. Posterior Urethral Valves

    Directory of Open Access Journals (Sweden)

    Steve J. Hodges

    2009-01-01

    Full Text Available The most common cause of lower urinary tract obstruction in male infants is posterior urethral valves. Although the incidence has remained stable, the neonatal mortality for this disorder has improved due to early diagnosis and intensive neonatal care, thanks in part to the widespread use of prenatal ultrasound evaluations. In fact, the most common reason for the diagnosis of posterior urethral valves presently is the evaluation of infants for prenatal hydronephrosis. Since these children are often diagnosed early, the urethral obstruction can be alleviated rapidly through catheter insertion and eventual surgery, and their metabolic derangements can be normalized without delay, avoiding preventable infant mortality. Of the children that survive, however, early diagnosis has not had much effect on their long-term prognosis, as 30% still develop renal insufficiency before adolescence. A better understanding of the exact cause of the congenital obstruction of the male posterior urethra, prevention of postnatal bladder and renal injury, and the development of safe methods to treat urethral obstruction prenatally (and thereby avoiding the bladder and renal damage due to obstructive uropathy are the goals for the care of children with posterior urethral valves[1].

  19. Anterior Urethral Valves

    Directory of Open Access Journals (Sweden)

    Vidyadhar P. Mali

    2006-07-01

    Full Text Available We studied the clinical presentation and management of four patients with anterior urethral valves; a rare cause of urethral obstruction in male children. One patient presented antenatally with oligohydramnios, bilateral hydronephrosis and bladder thickening suggestive of an infravesical obstruction. Two other patients presented postnatally at 1 and 2 years of age, respectively, with poor stream of urine since birth. The fourth patient presented at 9 years with frequency and dysuria. Diagnosis was established on either micturating cystourethrogram (MCU (in 2 or on cystoscopy (in 2. All patients had cystoscopic ablation of the valves. One patient developed a postablation stricture that was resected with an end-to-end urethroplasty. He had an associated bilateral vesicoureteric junction (VUJ obstruction for which a bilateral ureteric reimplantation was done at the same time. On long-term follow-up, all patients demonstrated a good stream of urine. The renal function is normal. Patients are continent and free of urinary infections. Anterior urethral valves are rare obstructive lesions in male children. The degree of obstruction is variable, and so they may present with mild micturition difficulty or severe obstruction with hydroureteronephrosis and renal impairment. Hence, it is important to evaluate the anterior urethra in any male child with suspected infravesical obstruction. The diagnosis is established by MCU or cystoscopy and the treatment is always surgical, either a transurethral ablation or an open resection. The long-term prognosis is good.

  20. Steam Turbine Control Valve Stiction Effect on Power System Stability

    International Nuclear Information System (INIS)

    Halimi, B.

    2010-01-01

    One of the most important problems in power system dynamic stability is low frequency oscillations. This kind of oscillation has significant effects on the stability and security of the power system. In some previous papers, a fact was introduced that a steam pressure continuous fluctuation in turbine steam inlet pipeline may lead to a kind of low frequency oscillation of power systems. Generally, in a power generation plant, steam turbine system composes of some main components, i.e. a boiler or steam generator, stop valves, control valves and turbines that are connected by piping. In the conventional system, the turbine system is composed with a lot of stop and control valves. The steam is provided by a boiler or steam generator. In an abnormal case, the stop valve shuts of the steal flow to the turbine. The steam flow to the turbine is regulated by controlling the control valves. The control valves are provided to regulate the flow of steam to the turbine for starting, increasing or decreasing the power, and also maintaining speed control with the turbine governor system. Unfortunately, the control valve has inherent static friction (stiction) nonlinearity characteristics. Industrial surveys indicated that about 20-30% of all control loops oscillate due to valve problem caused by this nonlinear characteristic. In this paper, steam turbine control valve stiction effect on power system oscillation is presented. To analyze the stiction characteristic effect, firstly a model of control valve and its stiction characteristic are derived by using Newton's laws. A complete tandem steam prime mover, including a speed governing system, a four-stage steam turbine, and a shaft with up to for masses is adopted to analyze the performance of the steam turbine. The governor system consists of some important parts, i.e. a proportional controller, speed relay, control valve with its stiction characteristic, and stem lift position of control valve controller. The steam turbine has

  1. Lessons from the Fukushima nuclear power accident

    International Nuclear Information System (INIS)

    Hatamura, Yotaro

    2013-01-01

    Through the investigation of the Fukushima Nuclear Power Accident as the chairman of the related Government's Committee, many things had been considered. Essence of the accident could be not only what occurred in the Fukushima nuclear power station, but also dispersed radioactive materials forced many residents to move and not to be returned. Such events as indication errors of water level meter occurring in severe accident could no be thought and remote mechanical operation of valves under high radiation environment were not prepared. Contamination by radioactive clouds caused the evacuation of residents for a long period. Lessons learned from the accident were described such as; (1) the verification of the road to failure connecting selected accident sequence and road to success with another supposed choice, (2) considering what might occur and then what should be needed on the contrary, (3) nuclear power, if should be continued, should be used with the premise of its hazards, and (4) advise to nuclear engineer for adequate information dissemination and technical explanation to the public and keeping nuclear technologies alive. (T. Tanaka)

  2. Overview of Prevention for Water Hammer by Check Valve Action in Nuclear Reactor

    International Nuclear Information System (INIS)

    Kim, Dayong; Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonhoon

    2016-01-01

    Water hammer can cause serious damage to pumping system and unexpected system pressure rise in the pipeline. In nuclear reactor, water hammer can influence on the integrity of safety related system. Water hammer in nuclear reactor have been caused by voiding in normally water-filled lines, steam condensation line containing both steam and water, as well as by rapid check valve action. Therefore, this study focuses on the water hammer by check valve among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. This study focuses on the water hammer by check valve action among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. If the inadvertent pump trip or pipe rupture in high velocity and pressure pipe is predicted, the fast response check valve such as tiled disc, dual disc and nozzle check valve should be installed in the system. If the inadvertent pump trip or pipe rupture in very high velocity and pressure pipe and excessively large revered flow velocity are predicted, the very slowly closing check valve such as controlled closure check valve should be installed in the system

  3. Overview of Prevention for Water Hammer by Check Valve Action in Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dayong; Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonhoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Water hammer can cause serious damage to pumping system and unexpected system pressure rise in the pipeline. In nuclear reactor, water hammer can influence on the integrity of safety related system. Water hammer in nuclear reactor have been caused by voiding in normally water-filled lines, steam condensation line containing both steam and water, as well as by rapid check valve action. Therefore, this study focuses on the water hammer by check valve among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. This study focuses on the water hammer by check valve action among the sources of water hammer occurrence and suggests proper methodology for check valve type selection against water hammer. If the inadvertent pump trip or pipe rupture in high velocity and pressure pipe is predicted, the fast response check valve such as tiled disc, dual disc and nozzle check valve should be installed in the system. If the inadvertent pump trip or pipe rupture in very high velocity and pressure pipe and excessively large revered flow velocity are predicted, the very slowly closing check valve such as controlled closure check valve should be installed in the system.

  4. Fracturing mechanics before valve-in-valve therapy of small aortic bioprosthetic heart valves.

    Science.gov (United States)

    Johansen, Peter; Engholt, Henrik; Tang, Mariann; Nybo, Rasmus F; Rasmussen, Per D; Nielsen-Kudsk, Jens Erik

    2017-10-13

    Patients with degraded bioprosthetic heart valves (BHV) who are not candidates for valve replacement may benefit from transcatheter valve-in-valve (VIV) therapy. However, in smaller-sized surgical BHV the resultant orifice may become too narrow. To overcome this, the valve frame can be fractured by a high-pressure balloon prior to VIV. However, knowledge on fracture pressures and mechanics are prerequisites. The aim of this study was to identify the fracture pressures needed in BHV, and to describe the fracture mechanics. Commonly used BHV of small sizes were mounted on a high-pressure balloon situated in a biplane fluoroscopic system with a high-speed camera. The instant of fracture was captured along with the balloon pressure. The valves were inspected for material protrusion and later dissected for fracture zone investigation and description. The valves with a polymer frame fractured at a lower pressure (8-10 atm) than those with a metal stent (19-26 atm). None of the fractured valves had elements protruding. VIV procedures in small-sized BHV may be performed after prior fracture of the valve frame by high-pressure balloon dilatation. This study provides tentative guidelines for expected balloon sizes and pressures for valve fracturing.

  5. Fast-Valving of Large Steam Turbine Units as a Means of Power System Security Enhancement

    Directory of Open Access Journals (Sweden)

    Bogdan Sobczak

    2014-03-01

    Full Text Available Fast-valving assists in maintaining system stability following a severe transmission system fault by reducing the turbine mechanical power. Fast-valving consists in rapid closing and opening of steam valves in an adequate manner to reduce the generator accelerating power following the recognition of a severe fault. FV can be an effective and economical method of meeting the performance requirements of a power system in the presence of an increase in wind and solar generation in the power system, newly connected large thermal units and delaying of building new transmission lines. The principle of fast-valving and advantages of applying this technique in large steam turbine units was presented in the paper. Effectiveness of fast-valving in enhancing the stability of the Polish Power Grid was analyzed. The feasibility study of fast-valving application in the 560 MW unit in Kozienice Power Station (EW SA was discussed.

  6. Accident information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

  7. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  8. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  9. Flow analysis by using solenoid valves for As(III determination in natural waters by an on-line separation and pre-concentration system coupled to a tungsten coil atomizer

    Directory of Open Access Journals (Sweden)

    José Y. Neira

    2005-03-01

    Full Text Available A flow system coupled to a tungsten coil atomizer in an atomic absorption spectrometer (TCA-AAS was developed for As(III determination in waters, by extraction with sodium diethyldithiocarbamate (NaDDTC as complexing agent, and by sorption of the As(III-DDTC complex in a micro-column filled with 5 mg C18 reversed phase (10 µL dry sorbent, followed by elution with ethanol. A complete pre-concentration/elution cycle took 208 s, with 30 s sample load time (1.7 mL and 4 s elution time (71 µL. The interface and software for the synchronous control of two peristaltic pumps (RUN/ STOP, an autosampler arm, seven solenoid valves, one injection valve, the electrothermal atomizer and the spectrometer Read function were constructed. The system was characterized and validated by analytical recovery studies performed both in synthetic solutions and in natural waters. Using a 30 s pre-concentration period, the working curve was linear between 0.25 and 6.0 µg L-1 (r = 0.9976, the retention efficiency was 94±1% (6.0 µg L-1, and the pre-concentration coefficient was 28.9. The characteristic mass was 58 pg, the mean repeatability (expressed as the variation coefficient was 3.4% (n=5, the detection limit was 0.058 µg L-1 (4.1 pg in 71 µL of eluate injected into the coil, and the mean analytical recovery in natural waters was 92.6 ± 9.5 % (n=15. The procedure is simple, economic, less prone to sample loss and contamination and the useful lifetime of the micro-column was between 200-300 pre-concentration cycles.

  10. Automatic fire hydrant valve development

    International Nuclear Information System (INIS)

    Drumheller, K.

    1976-01-01

    The development of a remotely-controlled valve to operate a fire hydrant is described. Assembled from off-the-shelf components, the prototype illustrates that a valve light enough to be handled by one man is possible. However, it does not have the ruggedness or reliability needed for actual fire-fighting operations. Preliminary testing by City of Tacoma fire department personnel indicates that the valve may indeed contribute significantly to fire-fighting efficiency

  11. [Ahmed valve in glaucoma surgery].

    Science.gov (United States)

    Bikbov, M M; Khusnitdinov, I I

    This is a review on Ahmed valve application in glaucoma surgery. It contains, in particular, data on the Ahmed valve efficiency, results of experimental and histological studies of filtering bleb encapsulation, examines the use of antimetabolites and anti-VEGF agents, and discusses implantation techniques. The current appraisal of antimetabolites delivery systems integrated into the Ahmed valve is presented. Various complications encountered in practice and preventive measures are also covered.

  12. Nonlinear transient dynamic response of pressure relief valves for a negative containment system

    International Nuclear Information System (INIS)

    Aziz, T.S.; Duff, C.G.; Tang, J.H.K.

    1979-01-01

    The response of the piston for the postulated simultaneous effect of pressure and an earthquake is obtained for different parameters and accident conditions. Response quantities such as accelerations, displacements, rotations, diaphragm forces as well as opening time during a design basis earthquake are obtained. The results of the different analyses, as related to the functional operability of the valves, are evaluated and discussed. (orig.)

  13. Alterations in the evaporation and discharge calculations for safety and relief valves in the Almod pressurizer

    International Nuclear Information System (INIS)

    Madeira, A.A.

    1986-01-01

    Models to estimate bubble rise velocity for evaporation, and critical mass flow for pressurizer relief and safety valves discharge calculation were implemented in ALMOD, a digital code developed to perform primary loop simulation of a PWR type during operational transients or accidents without loss of coolant. These models can be utilized alternatively, depending on the requirements for the analyzed transient condition. (Author) [pt

  14. 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

  15. Severe accident phenomena

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

    1995-02-01

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  16. Safety regulations regarding to accident monitoring and accident sampling at Russian NPPs with VVER type reactors

    International Nuclear Information System (INIS)

    Sharafutdinov, Rachet; Lankin, Michail; Kharitonova, Nataliya

    2014-01-01

    The paper describes a tendency by development of regulatory document requirements related to accident monitoring and accident sampling at Russia's NPPs. Lessons learned from the Fukushima Daiichi accident pointed at the importance and necessary to carry out an additional safety check at Russia's nuclear power plants in the preparedness for management of severe accidents at NPPs. Planned measures for improvement of severe accidents management include development and implementation of the accident instrumentation systems, providing, monitoring, management and storage of information in a severe accident conditions. The draft of Safety Guidelines <accident monitoring system of nuclear power plants with VVER reactors' prepared by Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) established the main criteria for accident monitoring instrumentation that can monitor relevant plant parameters in the reactor and inside containment during and after a severe accident in nuclear power plants. Development of these safety guidelines is in line with the recommendations of IAEA Action Plan on Nuclear Safety in response to the Fukushima Daiichi event and recommendations of the IAEA Nuclear Energy series Report <<Accident Monitoring Systems for Nuclear Power Plants' (Draft V 2.7). The paper presents the principles, which are used as the basis for selection of plant parameters for accident monitoring and for establishing of accident monitoring instrumentation. The recommendations to the accident sampling system capable to obtain the representative reactor coolant and containment air and fluid samples that support accurate analytical results for the parameters of interest are considered. The radiological and chemistry parameters to be monitored for primary coolant and sump and for containment air are specified. (author)

  17. Latest design of gate valves

    Energy Technology Data Exchange (ETDEWEB)

    Kurzhofer, U.; Stolte, J.; Weyand, M.

    1996-12-01

    Babcock Sempell, one of the most important valve manufacturers in Europe, has delivered valves for the nuclear power industry since the beginning of the peaceful application of nuclear power in the 1960s. The latest innovation by Babcock Sempell is a gate valve that meets all recent technical requirements of the nuclear power technology. At the moment in the United States, Germany, Sweden, and many other countries, motor-operated gate and globe valves are judged very critically. Besides the absolute control of the so-called {open_quotes}trip failure,{close_quotes} the integrity of all valve parts submitted to operational forces must be maintained. In case of failure of the limit and torque switches, all valve designs have been tested with respect to the quality of guidance of the gate. The guidances (i.e., guides) shall avoid a tilting of the gate during the closing procedure. The gate valve newly designed by Babcock Sempell fulfills all these characteristic criteria. In addition, the valve has cobalt-free seat hardfacing, the suitability of which has been proven by friction tests as well as full-scale blowdown tests at the GAP of Siemens in Karlstein, West Germany. Babcock Sempell was to deliver more than 30 gate valves of this type for 5 Swedish nuclear power stations by autumn 1995. In the presentation, the author will report on the testing performed, qualifications, and sizing criteria which led to the new technical design.

  18. Fluid mechanics of heart valves.

    Science.gov (United States)

    Yoganathan, Ajit P; He, Zhaoming; Casey Jones, S

    2004-01-01

    Valvular heart disease is a life-threatening disease that afflicts millions of people worldwide and leads to approximately 250,000 valve repairs and/or replacements each year. Malfunction of a native valve impairs its efficient fluid mechanic/hemodynamic performance. Artificial heart valves have been used since 1960 to replace diseased native valves and have saved millions of lives. Unfortunately, despite four decades of use, these devices are less than ideal and lead to many complications. Many of these complications/problems are directly related to the fluid mechanics associated with the various mechanical and bioprosthetic valve designs. This review focuses on the state-of-the-art experimental and computational fluid mechanics of native and prosthetic heart valves in current clinical use. The fluid dynamic performance characteristics of caged-ball, tilting-disc, bileaflet mechanical valves and porcine and pericardial stented and nonstented bioprostheic valves are reviewed. Other issues related to heart valve performance, such as biomaterials, solid mechanics, tissue mechanics, and durability, are not addressed in this review.

  19. Double-disc gate valve

    International Nuclear Information System (INIS)

    Wheatley, S.J.

    1979-01-01

    The invention relates to an improvement in a conventional double-disc gate valve having a vertically movable gate assembly including a wedge, spreaders slidably engaged therewith, a valve disc carried by the spreaders. When the gate assembly is lowered to a selected point in the valve casing, the valve discs are moved transversely outward to close inlet and outlet ports in the casing. The valve includes hold-down means for guiding the disc-and-spreader assemblies as they are moved transversely outward and inward. If such valves are operated at relatively high differential pressures, they sometimes jam during opening. Such jamming has been a problem for many years in gate valves used in gaseous diffusion plants for the separation of uranium isotopes. The invention is based on the finding that the above-mentioned jamming results when the outlet disc tilts about its horizontal axis in a certain way during opening of the valve. In accordance with the invention, tilting of the outlet disc is maintained at a tolerable value by providing the disc with a rigid downwardly extending member and by providing the casing with a stop for limiting inward arcuate movement of the member to a preselected value during opening of the valve

  20. Surge-damping vacuum valve

    International Nuclear Information System (INIS)

    Bullock, J.C.; Kelley, B.E.

    1977-01-01

    A valve for damping out flow surges in a vacuum system is described. The surge-damping mechanism consists of a slotted, spring-loaded disk adjacent to the valve's vacuum port (the flow passage to the vacuum roughing pump). Under flow surge conditions, the differential pressure forces the disk into a sealing engagement with the vacuum port, thereby restricting the gas flow path to narrow slots in the disk's periphery. The increased flow damps out the flow surge. When pressure is equalized on both sides of the valve, the spring load moves the disk away from the port to restore full flow conductance through the valve

  1. Supplement No. 79-01A to IE Bulletin No. 79-01: Environmental qualification of Class 1E equipment (Deficiencies in the environmental qualification of ASCO solenoid valves)

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    Recently, a noncompliance report under 10 CFR Part 21 was received by the NRC from the Henry Pratt Company, manufacturer of butterfly valves which are installed in the primary containment at the Three Mile Island Unit 2 Nuclear Station. These butterfly valves are used for purge and exhaust purposes and are required to operate during accident conditions. The report discusses the use of an unqualified solenoid valve for a safety-related valve function which requires operation under accident conditions. The solenoid valve in question is Catalogue No. HT-8331A45, manufactured by the Automatic Switch Company (ASCO) of Florham Park, New Jersey. This pilot valve is used to pilot control the pneumatic valve actuators which are installed on the containment ventilation butterfly valves at this facility. The deficiency in these solenoid valves identified in the Part 21 Report concerns the parts made of acetal plastic material. The acetal disc holder assembly and bottom plug in the pilot valve assembly are stated by ASCO to have a maximum service limit of 400,000 Rad integrated dosage and 200 degrees F temperature. According to ASCO, exposure of these acetal plastic parts to specified maximum environmental conditions may render the solenoid pilot valve inoperable which would cause the associated butterfly valve to malfunction

  2. Heart Valve Surgery Recovery and Follow Up

    Science.gov (United States)

    ... Guide: Understanding Your Heart Valve Problem | Spanish Symptom Tracker | Spanish Pre-surgery Checklist | Spanish What Is Heart ... Heart Valves • Heart Valve Problems and Causes • Risks, Signs and Symptoms • Accurate Diagnosis • Treatment Options • Recovery and ...

  3. Experimental and analytical studies on waterhammer generated by the closing of check valves

    International Nuclear Information System (INIS)

    Huet, J.L.; Garcia, J.L.; Coppolani, P.; Ziegler, B.

    1987-01-01

    A double-guillotine rupture on a water line upstream from a check valve generates a severe transient between the check valve and the pressure vessel on the downstream side. Successively following phenomena occur: - decrease then reversal of the flow, - closing of the check valve with impact of the plug on its seat, - waterhammer propagating in the pipe downstream from the check valve. The COMMISARIAT A L'ENERGIE ATOMIQUE (C.E.A.) FRAMATOME and ELECTRICITE DE FRANCE (E.D.F.) have undertaken a joint program in order to: - investigate the behavior uf the check valve in the event of a sudden closure, - evaluate the pressure and flow transient in the line. The program includes: - full scale tests in two loops, CLAUDIA (C.E.A.) and ECLAIR (E.D.F.), - analytical studies in order to qualify the calculation codes. This paper describes the experimental program and presents the analysis results for a benchmark test

  4. Aortic valve replacement with the Biocor PSB stentless xenograft.

    Science.gov (United States)

    Bertolini, P; Luciani, G B; Vecchi, B; Pugliese, P; Mazzucco, A

    1998-08-01

    The midterm clinical results after aortic valve replacement with the Biocor PSB stentless xenograft on all patients operated between October 1992 and October 1996 were reviewed. One hundred six patients, aged 70+/-6 years, had aortic valve replacement for aortic stenosis (67%), regurgitation (11%), or both (22%). Associated procedures were done in 49 patients (46%), including coronary artery bypass in 30 patients, mitral valve repair/replacement in 16, and ascending aorta replacement in 5 patients. Aortic cross-clamp and cardiopulmonary bypass times were 96+/-24 and 129+/-31 minutes, respectively. There were 3 (3%) early deaths due to low output (2 patients) and cerebrovascular accident (1 patient). Follow-up of survivors ranged from 6 to 66 months (mean, 39+/-14 months). Survival was 94%+/-2% and 90%+/-3% at 1 and 5 years. There were 5 late deaths due to cardiac cause (2), cancer (2), and pulmonary embolism (1 patient). No patient had structural valve deterioration, whereas 100% and 95%+/-3% were free from valve-related events at 1 and 5 years. There were two reoperations due to narrowing of the left coronary ostium and endocarditis, with an actuarial freedom from reoperation of 99%+/-1% and 98+/-1% at 1 and 5 years, respectively. Functional results demonstrated a mean peak transprosthetic gradient of 16+/-12 mm Hg, with only 1 patient (1%) with a 55 mm Hg gradient. No cases of valve regurgitation greater than mild were recorded at follow-up. Assessment of New York Heart Association functional class demonstrated a significant improvement (2.9+/-0.6 versus 1.4+/-0.7; p=0.01). All patients were free from anticoagulation. Aortic valve replacement using the Biocor PSB stentless xenograft offers excellent midterm survival, negligible valve deterioration, and a very low rate of valve-related events, which are comparable to estimates reported with other models of stentless xenografts and currently available stented xenografts. Hemodynamic performance is favorable and

  5. Analysis of traffic accidents in Romania, 2009.

    Science.gov (United States)

    Călinoiu, Geovana; Minca, Dana Galieta; Furtunescu, Florentina Ligia

    2012-01-01

    This paper aimed to underline the main consequences of traffic accidents in Romania 2009 and their associated causes or circumstances. We identified some problematic geographic areas, some critical months or moments of the day and also the most frequent causes; all these should become targets for the future planning. The current analysis provides some priority criteria for public health interventions. So, the future national road safety strategy should be in line with the EU objectives, but also with the national priorities. Romania is far away from the average EU target for 2010 of halving the death by traffic accidents registered in 2001. To describe the circumstances and the consequences related to traffic accidents registered in Romania, for the year 2009. An ecological study was conducted. The traffic accidents circumstances were analyzed in terms of magnitude, geographic space, time and cause. The consequences were analyzed as affected people and damaged cars. A total of 28,627 traffic accidents were registered in Romania during the year 2009. 2,796 people were killed and 27,968 were hospitalized and 42,443 cars were damaged. 3 of 4 accidents were caused by violations on behalf of the car drivers. Most common violations in car drivers were excess of speed and priority violations (52.4%). Among the pedestrians, 7 of 10 accidents were caused by illegal crossing. A higher number of accidents occurred during the summer months and during the evening hours (from 5.00 pm till 8.00 pm). The traffic accidents represent a real public health problem in Romania and a serious burden for the health system. The gap between Romania and the other EU member states needs to be diminished in the next decade. In this purpose, the future national road safety strategy should be in line with the EU objectives, but also with the national priorities. Research is needed to understand the causes and the socio-economical impact of traffic accidents and to define appropriate national

  6. 241-AN-A valve pit manifold valves and position indication acceptance test procedure

    Energy Technology Data Exchange (ETDEWEB)

    VANDYKE, D.W.

    1999-08-25

    This document describes the method used to test design criteria for gear actuated ball valves installed in 241-AN-A Valve Pit located at 200E Tank Farms. The purpose of this procedure is to demonstrate the following: Equipment is properly installed, labeled, and documented on As-Built drawings; New Manifold Valves in the 241-AN-A Valve Pit are fully operable using the handwheel of the valve operators; New valve position indicators on the valve operators will show correct valve positions; New valve position switches will function properly; and New valve locking devices function properly.

  7. Development of integrated accident management assessment technology

    International Nuclear Information System (INIS)

    Jung, Won Dea; Ha, Jae Joo; Jin, Young Ho

    2002-04-01

    This project aims to develop critical technologies for accident management through securing evaluation frameworks and supporting tools, in order to enhance capabilities coping with severe accidents. For the research goal, firstly under the viewpoint of accident prevention, on-line risk monitoring system and the analysis framework for human error have been developed. Secondly, the training/supporting systems including the training simulator and the off-site risk evaluation system have been developed to enhance capabilities coping with severe accidents. Four kinds of research results have been obtained from this project. Firstly, the framework and taxonomy for human error analysis has been developed for accident management. As the second, the supporting system for accident managements has been developed. Using data that are obtained through the evaluation of off-site risk for Younggwang site, the risk database as well as the methodology for optimizing emergency responses has been constructed. As the third, a training support system, SAMAT, has been developed, which can be used as a training simulator for severe accident management. Finally, on-line risk monitoring system, DynaRM, has been developed for Ulchin 3 and 4 unit

  8. Aortic valve replacement

    DEFF Research Database (Denmark)

    Kapetanakis, Emmanouil I; Athanasiou, Thanos; Mestres, Carlos A

    2008-01-01

    mortality were collected. Group analysis by patient geographic distribution and by annular diameter of the prosthesis utilized was conducted. Patients with a manufacturer's labeled prosthesis size > or = 21 mm were assigned to the 'large' aortic size subset, while those with a prosthesis size ... differences in the distribution of either gender or BSA. In the multivariable model, south European patients were seven times more likely to receive a smaller-sized aortic valve (OR = 6.5, 95% CI = 4.82-8.83, p

  9. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.; Thykier-Nielsen, S.; Walmod-Larsen, O.

    1986-08-01

    This report was commissioned by the Swedish State Power Board, who wanted a method for calculation of radiation doses in the surroundings of nuclear power plants caused by severe accidents. The PLUCON4 code were used for the calculations. A TC-SV-accident at Ringhals 1 wer chosen as example. A transient without shutdown leads to core meltdown through the reactor vessel. The pressure peak at the moment of vessel failure opens a safety valve in the dry well. Meteorolgical data for two years from the Ringhals meteorological tower were analysed to find representative weather situations. As typical weather were chosen Pasquill D with wind speed 8 m/s, and as extreme weather were chosen Pasquill F with wind speed 4.8 m/s. (author)

  10. BNGS B valve packing program

    International Nuclear Information System (INIS)

    Cumming, D.

    1995-01-01

    The Bruce B Valve Packing Program began in 1987. The early history and development were presented at the 1992 International CANDU Maintenance conference. This presentation covers the evolution of the Bruce B Valve Packing Program over the period 1992 to 1995. (author)

  11. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

    Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

    2014-01-01

    The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

  12. Experience with valves for PHWR reactors

    International Nuclear Information System (INIS)

    Narayan, K.; Mhetre, S.G.

    1977-01-01

    Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

  13. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  14. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  15. A cliff edge evaluation for CANDU-6 beyond design basis accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.M.; Kho, D.W., E-mail: wolsong@khnp.co.kr [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Yi, S.D.; Kang, S.H.; Kim, S.R. [Nuclear Engineering Service and Solution Co., Ltd., Daejeon (Korea, Republic of)

    2015-07-01

    The condition of nuclear power plant in the event of station black out (SBO) accompanying large-scale natural disaster exceeding design basis accident (DBA) was evaluated. Additional scenarios were added to the evaluation to review capability of the plant to endure different conditions with different actions. The analysis resulted that the key action required from the operator was to ensure the opening of main steam safety valves (MSSVs) in the secondary side and of motor-operated valves for high pressure injection of Emergency Core Cooling System (HPECCS) to mitigate accidents or extend the cliff edge. (author)

  16. New technology for accident prevention

    Energy Technology Data Exchange (ETDEWEB)

    Byne, P. [Shiftwork Solutions, Vancouver, BC (Canada)

    2006-07-01

    This power point presentation examined the effects of fatigue in the workplace and presented 3 technologies designed to prevent or monitor fatigue. The relationship between mental fatigue, circadian rhythms and cognitive performance was explored. Details of vigilance related degradations in the workplace were presented, as well as data on fatigue-related accidents and a time-line of meter-reading errors. It was noted that the direct cause of the Exxon Valdez disaster was sleep deprivation. Fatigue related accidents during the Gulf War were reviewed. The effects of fatigue on workplace performance include impaired logical reasoning and decision-making; impaired vigilance and attention; slowed mental operations; loss of situational awareness; slowed reaction time; and short cuts and lapses in optional or self-paced behaviours. New technologies to prevent fatigue-related accidents include (1) the driver fatigue monitor, an infra-red camera and computer that tracks a driver's slow eye-lid closures to prevent fatigue related accidents; (2) a fatigue avoidance scheduling tool (FAST) which collects actigraphs of sleep activity; and (3) SAFTE, a sleep, activity, fatigue and effectiveness model. refs., tabs., figs.

  17. Prosthetic valve endocarditis after transcatheter aortic valve implantation

    DEFF Research Database (Denmark)

    Olsen, Niels Thue; De Backer, Ole; Thyregod, Hans G H

    2015-01-01

    BACKGROUND: Transcatheter aortic valve implantation (TAVI) is an advancing mode of treatment for inoperable or high-risk patients with aortic stenosis. Prosthetic valve endocarditis (PVE) after TAVI is a serious complication, but only limited data exist on its incidence, outcome, and procedural......%) were treated conservatively and 1 with surgery. Four patients (22%) died from endocarditis or complications to treatment, 2 of those (11%) during initial hospitalization for PVE. An increased risk of TAVI-PVE was seen in patients with low implanted valve position (hazard ratio, 2.8 [1.1-7.2]), moderate...

  18. Nuclear accident dosimetry

    International Nuclear Information System (INIS)

    1982-01-01

    The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

  19. Nuclear accident dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-12-31

    The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

  20. Cavitation guide for control valves

    Energy Technology Data Exchange (ETDEWEB)

    Tullis, J.P. [Tullis Engineering Consultants, Logan, UT (United States)

    1993-04-01

    This guide teaches the basic fundamentals of cavitation to provide the reader with an understanding of what causes cavitation, when it occurs, and the potential problems cavitation can cause to a valve and piping system. The document provides guidelines for understanding how to reduce the cavitation and/or select control valves for a cavitating system. The guide provides a method for predicting the cavitation intensity of control valves, and how the effect of cavitation on a system will vary with valve type, valve function, valve size, operating pressure, duration of operation and details of the piping installation. The guide defines six cavitation limits identifying cavitation intensities ranging from inception to the maximum intensity possible. The intensity of the cavitation at each limit Is described, including a brief discussion of how each level of cavitation influences the valve and system. Examples are included to demonstrate how to apply the method, including making both size and pressure scale effects corrections. Methods of controlling cavitation are discussed providing information on various techniques which can be used to design a new system or modify an existing one so it can operate at a desired level of cavitation.

  1. Cavitation guide for control valves

    International Nuclear Information System (INIS)

    Tullis, J.P.

    1993-04-01

    This guide teaches the basic fundamentals of cavitation to provide the reader with an understanding of what causes cavitation, when it occurs, and the potential problems cavitation can cause to a valve and piping system. The document provides guidelines for understanding how to reduce the cavitation and/or select control valves for a cavitating system. The guide provides a method for predicting the cavitation intensity of control valves, and how the effect of cavitation on a system will vary with valve type, valve function, valve size, operating pressure, duration of operation and details of the piping installation. The guide defines six cavitation limits identifying cavitation intensities ranging from inception to the maximum intensity possible. The intensity of the cavitation at each limit Is described, including a brief discussion of how each level of cavitation influences the valve and system. Examples are included to demonstrate how to apply the method, including making both size and pressure scale effects corrections. Methods of controlling cavitation are discussed providing information on various techniques which can be used to design a new system or modify an existing one so it can operate at a desired level of cavitation

  2. Characteristic analysis of servo valve

    International Nuclear Information System (INIS)

    Ko, J. H.; Ryu, D. R.; Lee, J. H.; Kim, Y. S.; Na, J. C.; Kim, D. S.

    2008-01-01

    Electro-pneumatic servo valve is an electro-mechanical device which converts electric signals into a proper pneumatic flow rate or pressure. In order to improve the overall performance of pneumatic servo systems, electro-pneumatic servo valves are required, which have fast dynamic characteristics, no air leakage at a null point, and can be fabricated at a low-cost. The first objective of this research is to design and to fabricate a new electro-pneumatic servo valve which satisfies the above-mentioned requirements. In order to design the mechanism of the servo valve optimally, the flow inside the valve depending upon the position of spool was analyzed variously, and on the basis of such analysis results, the valve mechanism, which was formed by combination of the spool and the sleeve, was designed and manufactured. And a tester for conducting an overall performance test was designed and manufactured, and as a result of conducting the flow rate test, the pressure test and the frequency test on the developed pneumatic servo valve

  3. Developments in mechanical heart valve prosthesis

    Indian Academy of Sciences (India)

    Artificial heart valves are engineered devices used for replacing diseased or damaged natural valves of the heart. Most commonly used for replacement are mechanical heart valves and biological valves. This paper briefly outlines the evolution, designs employed, materials being used,. and important factors that affect the ...

  4. Butterfly valve torque prediction methodology

    International Nuclear Information System (INIS)

    Eldiwany, B.H.; Sharma, V.; Kalsi, M.S.; Wolfe, K.

    1994-01-01

    As part of the Motor-Operated Valve (MOV) Performance Prediction Program, the Electric Power Research Institute has sponsored the development of methodologies for predicting thrust and torque requirements of gate, globe, and butterfly MOVs. This paper presents the methodology that will be used by utilities to calculate the dynamic torque requirements for butterfly valves. The total dynamic torque at any disc position is the sum of the hydrodynamic torque, bearing torque (which is induced by the hydrodynamic force), as well as other small torque components (such as packing torque). The hydrodynamic torque on the valve disc, caused by the fluid flow through the valve, depends on the disc angle, flow velocity, upstream flow disturbances, disc shape, and the disc aspect ratio. The butterfly valve model provides sets of nondimensional flow and torque coefficients that can be used to predict flow rate and hydrodynamic torque throughout the disc stroke and to calculate the required actuation torque and the maximum transmitted torque throughout the opening and closing stroke. The scope of the model includes symmetric and nonsymmetric discs of different shapes and aspects ratios in compressible and incompressible fluid applications under both choked and nonchoked flow conditions. The model features were validated against test data from a comprehensive flowloop and in situ test program. These tests were designed to systematically address the effect of the following parameters on the required torque: valve size, disc shapes and disc aspect ratios, upstream elbow orientation and its proximity, and flow conditions. The applicability of the nondimensional coefficients to valves of different sizes was validated by performing tests on 42-in. valve and a precisely scaled 6-in. model. The butterfly valve model torque predictions were found to bound test data from the flow-loop and in situ testing, as shown in the examples provided in this paper

  5. Analysis of Fukushima Daiichi Accident Using HFACS

    International Nuclear Information System (INIS)

    Mohamed, Saeed Almheiri

    2013-01-01

    The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO 1 and NISA 2 that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident

  6. Analysis of Fukushima Daiichi Accident Using HFACS

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Saeed Almheiri [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

    2013-10-15

    The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO{sup 1} and NISA{sup 2} that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident.

  7. Sequential injection/bead injection lab-on-valve schemes for on-line solid phase extraction and preconcentration of ultra-trace levels of heavy metals with determination by ETAAS and ICPMS

    DEFF Research Database (Denmark)

    Wang, Jianhua; Hansen, Elo Harald; Miró, Manuel

    2003-01-01

    are focused on the applications of SI-BI-LOV protocols for on-line microcolumn based solid phase extraction of ultra-trace levels of heavy metals, employing the so-called renewable surface separation and preconcentration manipulatory scheme. Two types of sorbents have been employed as packing material...

  8. Thermostatic Radiator Valve Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Dentz, J. [Advanced Residential Integrated Energy Solutions Collaborative (ARIES), New York, NY (United States); Ansanelli, E. [Advanced Residential Integrated Energy Solutions Collaborative (ARIES), New York, NY (United States)

    2015-01-01

    A large stock of multifamily buildings in the Northeast and Midwest are heated by steam distribution systems. Losses from these systems are typically high and a significant number of apartments are overheated much of the time. Thermostatically controlled radiator valves (TRVs) are one potential strategy to combat this problem, but have not been widely accepted by the residential retrofit market. In this project, the ARIES team sought to better understand the current usage of TRVs by key market players in steam and hot water heating and to conduct limited experiments on the effectiveness of new and old TRVs as a means of controlling space temperatures and reducing heating fuel consumption. The project included a survey of industry professionals, a field experiment comparing old and new TRVs, and cost-benefit modeling analysis using BEopt™ (Building Energy Optimization software).

  9. Several accidents about ERHRS of CEFR

    International Nuclear Information System (INIS)

    Zhang, D.

    2000-01-01

    An analysis of about several unusual accidents about Emergency Residual Heat Removal System (ERHRS) of China Experiment Fast Reactor (CEFR) is presented. CEFR is a pool-type sodium-cooled fast reactor. The ERHRS of this reactor is designed in passive principle, which enhances the interior reliability of CEFR. It consists of two sets of independent channels. Each channel is comprised of decay heat exchanger (DHX), intermediate circuit, sodium-air heat exchanger (AHX) and related auxiliary system. Both DHX are located in the hot pool of the main vessel directly, which is used to cool the hot sodium. The whole set of ERHRS is completely passive except the ventilation valves of AHX. But, as a very important set of engineered safety features which is the final way to remove the heat from the reactor core, it is necessary to pay attention to all of the possibilities that may reduce this ability. Several accidents are analyzed including when the ventilation valves couldn't be opened, when only one set of ERHRS could work and so on. The calculation results show that the ERHRS can keep the reactor in a safety status. Even though it is, experiments are still necessary in the view of engineering. (author)

  10. Transcatheter Mitral Valve-in-Ring Implantation

    LENUS (Irish Health Repository)

    Tanner, RE

    2018-05-01

    Failed surgical mitral valve repair using an annuloplasty ring has traditionally been treated with surgical valve replacement or repair1. For patients at high risk for repeat open heart surgery, placement of a trans-catheter aortic valve (i.e., TAVI valve) within the mitral ring (i.e., Mitral-Valve-in-Ring, MViR) has emerged as a novel alternative treatment strategy2-5 . We describe our experience of a failed mitral valve repair that was successfully treated with a TAVI valve delivered via the trans-septal approach, and summarise the data relating to this emerging treatment strategy.

  11. Valve leakage inspection testing and maintenance process

    International Nuclear Information System (INIS)

    Aikin, J.A.; Reinwald, J.W.; Kittmer, C.A.

    1991-01-01

    In valve maintenance, packing rings that prevent leakage along the valve stem must periodically be replaced, either during routine maintenance or to correct a leak or valve malfunction. Tools and procedures currently in use for valve packing removal and inspection are generally of limited value due to various access and application problems. A process has been developed by AECL Research that addresses these problems. The process, using incompressible fluid pressure, quickly and efficiently confirms the integrity of the valve backseat, extracts hard-to-remove valve packing sets, and verifies the leak tightness of the repacked valve

  12. Transcatheter, valve-in-valve transapical aortic and mitral valve implantation, in a high risk patient with aortic and mitral prosthetic valve stenoses

    Directory of Open Access Journals (Sweden)

    Harish Ramakrishna

    2015-01-01

    Full Text Available Transcatheter valve implantation continues to grow worldwide and has been used principally for the nonsurgical management of native aortic valvular disease-as a potentially less invasive method of valve replacement in high-risk and inoperable patients with severe aortic valve stenosis. Given the burden of valvular heart disease in the general population and the increasing numbers of patients who have had previous valve operations, we are now seeing a growing number of high-risk patients presenting with prosthetic valve stenosis, who are not potential surgical candidates. For this high-risk subset transcatheter valve delivery may be the only option. Here, we present an inoperable patient with severe, prosthetic valve aortic and mitral stenosis who was successfully treated with a trans catheter based approach, with a valve-in-valve implantation procedure of both aortic and mitral valves.

  13. Supervisor's accident investigation handbook

    International Nuclear Information System (INIS)

    1980-02-01

    This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

  14. Anatomic, histopathologic, and echocardiographic features in a dog with an atypical pulmonary valve stenosis with a fibrous band of tissue and a patent ductus arteriosus.

    Science.gov (United States)

    Yoon, Hakyoung; Kim, Jaehwan; Nahm, Sang-Soep; Eom, Kidong

    2017-07-11

    Congenital pulmonary valve stenosis and patent ductus arteriosus are common congenital heart defects in dogs. However, concurrence of atypical pulmonary valve stenosis and patent ductus arteriosus is uncommon. This report describes the anatomic, histopathologic, and echocardiographic features in a dog with concomitant pulmonary valve stenosis and patent ductus arteriosus with atypical pulmonary valve dysplasia that included a fibrous band of tissue. A 1.5-year-old intact female Chihuahua dog weighing 3.3 kg presented with a continuous grade VI cardiac murmur, poor exercise tolerance, and an intermittent cough. Echocardiography indicated pulmonary valve stenosis, a thickened dysplastic valve without annular hypoplasia, and a type IIA patent ductus arteriosus. The pulmonary valve was thick line-shaped in systole and dome-shaped towards the right ventricular outflow tract in diastole. The dog suffered a fatal cardiac arrest during an attempted balloon pulmonary valvuloplasty. Necropsy revealed pulmonary valve dysplasia, commissural fusion, and incomplete opening and closing of the pulmonary valve because of a fibrous band of tissue causing adhesion between the right ventricular outflow tract and the dysplastic intermediate cusp of the valve. A fibrous band of tissue between the right ventricular outflow track and the pulmonary valve should be considered as a cause of pulmonary valve stenosis. Pulmonary valve stenosis and patent ductus arteriosus can have conflicting effects on diastolic and systolic dysfunction, respectively. Therefore, beta-blockers should always be used carefully, particularly in patients with a heart defect where there is concern about left ventricular systolic function.

  15. Anesthetic management for combined mitral valve replacement and aortic valve repair in a patient with osteogenesis imperfecta

    Directory of Open Access Journals (Sweden)

    Huang Jiapeng

    2011-01-01

    Full Text Available Osteogenesis imperfecta is a rare disorder of connective tissues and presents multiple challenges, including difficult airway, hyperthermia, coagulopathy and respiratory dysfunction, for anesthesiologists, especially during cardiac surgery. We present anesthetic management of a patient with osteogenesis impertecta during double valve surgery. Dexmedetomidine infusion minimized the risks of malignant hyperthermia. Glidescope and in-line stabilization facilitated endotracheal intubation and protected his oral structures and cervical spine. Transesophageal echocardiography (TEE diagnosed a flail A3 segment and redundant left coronary cusp causing mitral and aortic regurgitation. The mitral valve was replaced and the aortic valve repaired. Coagulopathy was corrected according to comprehensive coagulation analysis. Glidescope, dexmedetomidine, coagulation analysis and TEE could facilitate anesthetic management in these patients.

  16. Traumatic tricuspid valve insufficiency. Experience in thirteen patients.

    Science.gov (United States)

    van Son, J A; Danielson, G K; Schaff, H V; Miller, F A

    1994-11-01

    From 1964 through June 1993, thirteen patients with traumatic tricuspid insufficiency were treated surgically; all were male, and the ages ranged from 17 to 64 years (median 39 years). The condition was associated with blunt chest trauma in all patients: motor vehicle accidents in twelve and an explosion of a tank of compressed air in one. The median duration between trauma and operation was 17 years (range 1 month to 37 years). Preoperatively, six patients were in sinus rhythm and seven were in atrial fibrillation. At operation, the right ventricular function appeared moderately to severely depressed in twelve patients. In twelve patients, the anterior leaflet was flail because of chordal rupture (n = 9), rupture of anterior papillary muscle (n = 3), or tear in the anterior leaflet (n = 1). In one patient, the septal leaflet was missing and in another it was retracted and adherent to the ventricular septum. In five patients the tricuspid valve was repaired and in eight it was replaced. In seven patients in the latter group, the chordae, papillary muscles, and/or tricuspid valve leaflet(s) were found to be in a contracted and atrophic state, precluding repair. No early or late deaths occurred. At follow-up extending to 26 years (median 12 years), 12 patients are in New York Heart Association class I and one patient is in class II. Nine patients were in sinus rhythm and four were in atrial fibrillation. Although our experience indicates that good functional results can still be achieved many years after the onset of traumatic tricuspid valve insufficiency, earlier diagnosis and surgical treatment should increase the feasibility of tricuspid valve insufficiency, earlier diagnosis and surgical treatment should increase the feasibility of tricuspid valve repair, prevent progressive deterioration of right ventricular function, and increase the possibility of maintaining late sinus rhythm in a greater number of patients.

  17. Options for Heart Valve Replacement

    Science.gov (United States)

    ... Guide: Understanding Your Heart Valve Problem | Spanish Symptom Tracker | Spanish Pre-surgery Checklist | Spanish What Is Heart ... Cardiac Arrest: How Are They Different? 7 Warning Signs of a Heart Attack 8 Low Blood Pressure - ...

  18. Minimally invasive aortic valve replacement

    DEFF Research Database (Denmark)

    Foghsgaard, Signe; Schmidt, Thomas Andersen; Kjaergard, Henrik K

    2009-01-01

    In this descriptive prospective study, we evaluate the outcomes of surgery in 98 patients who were scheduled to undergo minimally invasive aortic valve replacement. These patients were compared with a group of 50 patients who underwent scheduled aortic valve replacement through a full sternotomy...... operations were completed as mini-sternotomies, 4 died later of noncardiac causes. The aortic cross-clamp and perfusion times were significantly different across all groups (P replacement...... is an excellent operation in selected patients, but its true advantages over conventional aortic valve replacement (other than a smaller scar) await evaluation by means of randomized clinical trial. The "extended mini-aortic valve replacement" operation, on the other hand, is a risky procedure that should...

  19. Embryological origin of the endocardium and derived valve progenitor cells: from developmental biology to stem cell-based valve repair.

    Science.gov (United States)

    Pucéat, Michel

    2013-04-01

    The cardiac valves are targets of both congenital and acquired diseases. The formation of valves during embryogenesis (i.e., valvulogenesis) originates from endocardial cells lining the myocardium. These cells undergo an endothelial-mesenchymal transition, proliferate and migrate within an extracellular matrix. This leads to the formation of bilateral cardiac cushions in both the atrioventricular canal and the outflow tract. The embryonic origin of both the endocardium and prospective valve cells is still elusive. Endocardial and myocardial lineages are segregated early during embryogenesis and such a cell fate decision can be recapitulated in vitro by embryonic stem cells (ESC). Besides genetically modified mice and ex vivo heart explants, ESCs provide a cellular model to study the early steps of valve development and might constitute a human therapeutic cell source for decellularized tissue-engineered valves. This article is part of a Special Issue entitled: Cardiomyocyte Biology: Cardiac Pathways of Differentiation, Metabolism and Contraction. Copyright © 2012 Elsevier B.V. All rights reserved.

  20. [Traumatic tricuspid valve insufficiency with right-to-left shunt: bridging using extracorporeal venovenous membrane oxygenation].

    Science.gov (United States)

    Weber, S U; Hammerstingl, C; Mellert, F; Baumgarten, G; Putensen, C; Knuefermann, P

    2012-01-01

    The case of a young male motor vehicle driver is reported who suffered multiple trauma in a car accident with pulmonary and cardiac contusions. In the course of severe pneumonia and traumatic tricuspid valve insufficiency a right-to-left shunt with refractory hypoxemia developed across a pre-existing atrial septal defect (ASD). The patient could be successfully treated by the combination of extracorporeal membrane oxygenation for bridging, interventional ASD occlusion and in the long-term by operative reconstruction of the tricuspid valve.

  1. A nuclear radiation actuated valve for a nuclear reactor

    International Nuclear Information System (INIS)

    Christiansen, D.W.; Schively, D.P.

    1983-01-01

    The valve has a first part (such as a valve rod with piston) and a second part (such as a valve tube surrounding the valve rod, with the valve tube having side slots surrounding the piston). Both valve parts have known nuclear radiation swelling characteristics which are different. The valve parts are positioned so that the valve's first and second parts create a valve orifice which changes in size due to the different nuclear radiation caused swelling of the valve's first part compared to the valve's second part. The valve may be used in a nuclear reactor's core coolant system. (author)

  2. An improved gate valve for critical applications in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kalsi, M.S.; Alvarez, P.D.; Wang, J.K.; Somagyi, D. [Kalsi Engineering, Inc., Sugar Land, TX (United States)] [and others

    1996-12-01

    U.S. Nuclear Regulatory Commission Generic Letters 89-10 for motor-operated valves (MOVs) and 95-07 for all power-operated valves document in detail the problems related to the performance of the safety-related valves in nuclear power plants. The problems relate to lack of reliable operation under design basis conditions including higher than anticipated stem thrust, unpredictable valve behavior, damage to the valve internals under blowdown/high flow conditions, significant degradation of performance when cycled under AP and flow, thermal binding, and pressure locking. This paper describes an improved motor-operated flexible wedge gate valve design, the GE Sentinel Valve, which is the outcome of a comprehensive and systematic development effort undertaken to resolve the issues identified in the NRC Generic Letters 89-10 and 95-07. The new design provides a reliable, long-term, low maintenance cost solution to the nuclear power industry. One of the key features incorporated in the disc permits the disc flexibility to be varied independently of the disc thickness (pressure boundary) dictated by the ASME Section III Pressure Vessel & Piping Code stress criteria. This feature allows the desired flexibility to be incorporated in the disc, thus eliminating thermal binding problems. A matrix of analyses was performed using finite element and computational fluid dynamics approaches to optimize design for stresses, flexibility, leak-tightness, fluid flow, and thermal effects. The design of the entire product line was based upon a consistent set of analyses and design rules which permit scaling to different valve sizes and pressure classes within the product line. The valve meets all of the ASME Section III Code design criteria and the N-Stamp requirements. The performance of the valve was validated by performing extensive separate effects and plant in-situ tests. This paper summarizes the key design features, analyses, and test results.

  3. An improved gate valve for critical applications in nuclear power plants

    International Nuclear Information System (INIS)

    Kalsi, M.S.; Alvarez, P.D.; Wang, J.K.; Somagyi, D.

    1996-01-01

    U.S. Nuclear Regulatory Commission Generic Letters 89-10 for motor-operated valves (MOVs) and 95-07 for all power-operated valves document in detail the problems related to the performance of the safety-related valves in nuclear power plants. The problems relate to lack of reliable operation under design basis conditions including higher than anticipated stem thrust, unpredictable valve behavior, damage to the valve internals under blowdown/high flow conditions, significant degradation of performance when cycled under AP and flow, thermal binding, and pressure locking. This paper describes an improved motor-operated flexible wedge gate valve design, the GE Sentinel Valve, which is the outcome of a comprehensive and systematic development effort undertaken to resolve the issues identified in the NRC Generic Letters 89-10 and 95-07. The new design provides a reliable, long-term, low maintenance cost solution to the nuclear power industry. One of the key features incorporated in the disc permits the disc flexibility to be varied independently of the disc thickness (pressure boundary) dictated by the ASME Section III Pressure Vessel ampersand Piping Code stress criteria. This feature allows the desired flexibility to be incorporated in the disc, thus eliminating thermal binding problems. A matrix of analyses was performed using finite element and computational fluid dynamics approaches to optimize design for stresses, flexibility, leak-tightness, fluid flow, and thermal effects. The design of the entire product line was based upon a consistent set of analyses and design rules which permit scaling to different valve sizes and pressure classes within the product line. The valve meets all of the ASME Section III Code design criteria and the N-Stamp requirements. The performance of the valve was validated by performing extensive separate effects and plant in-situ tests. This paper summarizes the key design features, analyses, and test results

  4. Imaging of Mitral Valve Prolapse: What Can We Learn from Imaging about the Mechanism of the Disease?

    Directory of Open Access Journals (Sweden)

    Ronen Durst

    2015-07-01

    Full Text Available Mitral valve prolapse (MVP is the most common mitral valve disorder affecting 2%–3% of the general population. Two histological forms for the disease exist: Myxomatous degeneration and fibroelastic disease. Pathological evidence suggests the disease is not confined solely to the valve tissue, and accumulation of proteoglycans and fibrotic tissue can be seen in the adjacent myocardium of MVP patients. MVP is diagnosed by demonstrating valve tissue passing the annular line into the left atrium during systole. In this review we will discuss the advantages and limitations of various imaging modalities in their MVP diagnosis ability as well as the potential for demonstrating extra associated valvular pathologies.

  5. Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

    International Nuclear Information System (INIS)

    Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo

    2016-01-01

    An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted

  6. Design of a High Power Robotic Manipulator for Emergency Response to the Nuclear Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jongwon; Bae, Yeong-Geol; Kim, Myoung Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    An accident in a nuclear facility causes a great social cost. To prevent an unexpected nuclear accident from spreading to the catastrophic disaster, emergency response action in early stage is required. However, high radiation environment has been proved as a challenging obstacle for human workers to access to the accident site and take an action in previous accident cases. Therefore, emergency response robotic technology to be used in a nuclear accident site instead of human workers are actively conducted in domestically and internationally. Robots in an accident situation are required to carry out a variety of tasks depend on the types and patterns of accidents. An emergency response usually includes removing of debris, make an access road to a certain place and handling valves. These tasks normally involve high payload handling. A small sized high power robotic manipulator can be an appropriate candidate to deal with a wide spectrum of tasks in an emergency situation. In this paper, we discuss about the design of a high power robotic manipulator, which is capable of handling high payloads for an initial response action to the nuclear facility accident. In this paper, we presented a small sized high power robotic manipulator design. Actuator types of manipulator was selected and mechanical structure was discussed. In the future, the servo valve and hydraulic pump systems will be determined. Furthermore, control algorithms and test bed experiments will be also conducted.

  7. Risk assessment of K basin twelve-inch drain valve failure from a postulated seismic initiating event

    International Nuclear Information System (INIS)

    MORGAN, R.G.

    1999-01-01

    The Spent Nuclear Fuel (SNF) Project will transfer metallic SNF from the Hanford 105 K-East and 105 K-West Basins to safe interim storage in the Canister Storage Building in the 200 Area. The initial basis for design, fabrication, installation, and operation of the fuel removal systems was that the basin leak rates which could result from a postulated accident condition would not be excessive relative to reasonable recovery operations. However, an additional potential K Basin water leak path is through the K Basin drain valves. Three twelve-inch drain valves are located in the main basin bays along the north wall. The sumps containing the valves are filled with concrete which covers the drain valve body. Visual observations suggest that only the valve's bonnet and stem are exposed above the basin concrete floor. It was recognized, however, that damage of the drain valve bonnet or stem during a seismic initiating event could provide a potential K Basin water leak path. The objectives of this activity are to: (1) evaluate the risk of damaging the three twelve-inch drain valves located along the north wall of the main basin from a seismic initiating event, and (2) determine the associated potential leak rate from a damaged valve

  8. Risk assessment of K basin twelve-inch drain valve failure from a postulated seismic initiating event

    Energy Technology Data Exchange (ETDEWEB)

    MORGAN, R.G.

    1999-04-06

    The Spent Nuclear Fuel (SNF) Project will transfer metallic SNF from the Hanford 105 K-East and 105 K-West Basins to safe interim storage in the Canister Storage Building in the 200 Area. The initial basis for design, fabrication, installation, and operation of the fuel removal systems was that the basin leak rates which could result from a postulated accident condition would not be excessive relative to reasonable recovery operations. However, an additional potential K Basin water leak path is through the K Basin drain valves. Three twelve-inch drain valves are located in the main basin bays along the north wall. The sumps containing the valves are filled with concrete which covers the drain valve body. Visual observations suggest that only the valve's bonnet and stem are exposed above the basin concrete floor. It was recognized, however, that damage of the drain valve bonnet or stem during a seismic initiating event could provide a potential K Basin water leak path. The objectives of this activity are to: (1) evaluate the risk of damaging the three twelve-inch drain valves located along the north wall of the main basin from a seismic initiating event, and (2) determine the associated potential leak rate from a damaged valve.

  9. Visualization of Traffic Accidents

    Science.gov (United States)

    Wang, Jie; Shen, Yuzhong; Khattak, Asad

    2010-01-01

    Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

  10. On-line ion exchange preconcentration in a sequential injection lab-on-valve microsystem incorporating a renewable column with ETAAS for the trace-level determination of bismuth in urine and river sediment

    DEFF Research Database (Denmark)

    Wang, Jianhua; Hansen, Elo Harald

    2001-01-01

    A sequential injection system for on-line ion-exchange separation and preconcentration of trace-level amounts of metal ions with ensuing detection by electrothermal atomic absorption spectrometry (ETAAS) is described. Based on the use of a renewable microcolumn incorporated within an integrated l.......3% for the determination of 2.0 mug/l Bi (n = 7). The procedure was validated by determination of bismuth in a certified reference material CRM 320 (river sediment), and by bismuth spike recoveries in two human urine samples....

  11. Small sodium valve design and operating experience

    International Nuclear Information System (INIS)

    Abramson, R.; Elie, X.; Vercasson, M.; Nedelec, J.

    1974-01-01

    Conventionally, valves for sodium pipes smaller than 125 mm in diameter are called ''small sodium valves''. However, this limit should rather be considered as the lower limit o ''large sodium valves''. In fact, both the largest sizes of small valves and the smallest of large valves can be found in the range of 125-300 mm in diameter. Thus what is said about small valves also applies, for a few valve types, above the 125 mm limit. Sodium valves are described here in a general manner, with no manufacturing details except when necessary for understanding valve behavior. Operating experience is pointed out wherever possible. Finally, some information is given about ongoing or proposed development plans. (U.S.)

  12. Design of the Modular Pneumatic Valve Terminal

    Directory of Open Access Journals (Sweden)

    Jakub E. TAKOSOGLU

    2015-11-01

    Full Text Available The paper presents design of the modular pneumatic valve terminal, which was made on the basis of the patent application No A1 402905 „A valve for controlling fluid power drives, specially for pneumatic actuators, and the control system for fluid power drives valves”. The authors describe a method of operation of the system with double-acting valve and 5/2 (five ways and two position valve. Functions of the valve, and an example of application of the valve terminal in the production process were presented. 3D solid models of all the components of the valve were made. The paper presents a complete 3D model of the valve in various configurations. Using CAD-embedded SOLIDWORKS Flow Simulation computational fluid dynamics CFD analysis was also carried out of compressed air flow in the ways of the valve elements

  13. Development of a smart type motor operated valve for nuclear power plants

    Science.gov (United States)

    Kim, Chang-Hwoi; Park, Joo-Hyun; Lee, Dong-young; Koo, In-Soo

    2005-12-01

    In this paper, the design concept of the smart type motor operator valve for nuclear power plant was described. The development objective of the smart valve is to achieve superior accuracy, long-term reliability, and ease of use. In this reasons, developed smart valve has fieldbus communication such as deviceNet and Profibus-DP, auto-tuning PID controller, self-diagnostics, and on-line calibration capabilities. And also, to achieve pressure, temperature, and flow control with internal PID controller, the pressure sensor and transmitter were included in this valve. And, temperature and flow signal acquisition port was prepared. The developed smart valve will be performed equipment qualification test such as environment, EMI/EMC, and vibration in Korea Test Lab. And, the valve performance is tested in a test loop which is located in Seoul National University Lab. To apply nuclear power plant, the software is being developed according to software life cycle. The developed software is verified by independent software V and V team. It is expected that the smart valve can be applied to an existing NPPs for replacing or to a new nuclear power plants. The design and fabrication of smart valve is now being processed.

  14. Development of an effective valve packing program

    Energy Technology Data Exchange (ETDEWEB)

    Hart, K.A.

    1996-12-01

    Current data now shows that graphite valve packing installed within the guidance of a controlled program produces not only reliable stem sealing but predictable running loads. By utilizing recent technological developments in valve performance monitoring for both MOV`s and AOV`s, valve packing performance can be enhanced while reducing maintenance costs. Once known, values are established for acceptable valve packing loads, the measurement of actual valve running loads via the current MOV/AOV diagnostic techniques can provide indication of future valve stem sealing problems, improper valve packing installation or identify the opportunity for valve packing program improvements. At times the full benefit of these advances in material and predictive technology remain under utilized due to simple past misconceptions associated with valve packing. This paper will explore the basis for these misconceptions, provide general insight into the current understanding of valve packing and demonstrate how with this new understanding and current valve diagnostic equipment the key aspects required to develop an effective, quality valve packing program fit together. The cost and operational benefits provided by this approach can be significant impact by the: elimination of periodic valve repacking, reduction of maintenance costs, benefits of leak-free valve operation, justification for reduced Post Maintenance Test Requirements, reduced radiation exposure, improved plant appearance.

  15. Soft valves in plants

    Science.gov (United States)

    Park, Keunhwan; Tixier, Aude; Christensen, Anneline; Arnbjerg-Nielsen, Sif; Zwieniecki, Maciej; Jensen, Kaare

    2017-11-01

    Water and minerals flow from plant roots to leaves in the xylem, an interconnected network of vascular conduits that spans the full length of the organism. When a plant is subjected to drought stress, air pockets can spread inside the xylem, threatening the survival of the plant. Many plants prevent propagation of air by using hydrophobic nano-membranes in the ``pit'' pores that link adjacent xylem cells. This adds considerable resistance to flow. Interestingly, torus-margo pit pores in conifers are open and offer less resistance. To prevent propagation of air, conifers use a soft gating mechanism, which relies on hydrodynamic interactions between the xylem liquid and the elastic pit. However, it is unknown exactly how it is able to combine the seemingly antagonist functions of high permeability and resistance to propagation of air. We conduct experiments on biomimetic pores to elucidate the flow regulation mechanism. The design of plant valves is compared to other natural systems and optimal strategies are discussed. This work was supported by a research Grant (13166) from VILLUM FONDEN.

  16. Annular flow diverter valve

    International Nuclear Information System (INIS)

    Rider, R.L.

    1980-01-01

    A valve is described for diverting flow from the center of two concentric tubes to the annulus between the tubes or, operating in the reverse direction, for mixing fluids from concentric tubes into a common tube and for controlling the volume ratio of said flow. It consists of a toroidal baffle disposed in sliding engagement with the interior of the inner tube downstream of a plurality of ports in the inner tube, a plurality of gates in sliding engagement with the interior of the inner tube attached to the baffle for movement therewith, a servomotor having a bullet-shaped plug on the downstream end thereof, and drive rods connecting the servomotor to the toroidal baffle. The sevomotor is adapted to move the baffle into mating engagement with the bullet-shaped plug and simultaneously move the gates away from the ports in the inner tube and to move the baffle away from the bullet-shaped plug and simultaneously move the gates to cover the ports in the inner tube

  17. Malfunction of a Heimlich flutter valve causing tension pneumothorax: case report of a rare complication

    Directory of Open Access Journals (Sweden)

    Braunstein Volker A

    2010-06-01

    Full Text Available Abstract Background Thoracic injuries play an important role in major trauma patients due to their high incidence and critical relevance. A serious consequence of thoracic trauma is pneumothorax, a condition that quickly can become life-threatening and requires immediate treatment. Decompression is the state of the art for treating tension pneumothorax. There are many different methods of decompression using different techniques, devices, valves and drainage systems. Referring to our case report we would like to discuss the utilization of these devices. Case presentation We report of a patient suffering from tension pneumothorax despite insertion of a chest drain at the accident scene. The decompression was by tube thoracostomy which was connected to a Heimlich flutter valve. During air transportation the patient suffered from cardiorespiratory arrest with asystole and was admitted to the trauma room undergoing manual chest compressions. The initial chest film showed a persisting tension pneumothorax, despite the chest tube that had been correctly placed and connected properly to the Heimlich valve. We assume that the Heimlich valve leaves did not open up and thus tension pneumothorax was not released. Conclusion We would like to raise awareness to the fact that if a Heimlich flutter valve is applied in the pre-hospital setting it should be used with caution. Failure in this type of valve may lead to recurrent tension pneumothorax.

  18. Bioprosthetic Valve Fracture During Valve-in-valve TAVR: Bench to Bedside.

    Science.gov (United States)

    Saxon, John T; Allen, Keith B; Cohen, David J; Chhatriwalla, Adnan K

    2018-01-01

    Valve-in-valve (VIV) transcatheter aortic valve replacement (TAVR) has been established as a safe and effective means of treating failed surgical bioprosthetic valves (BPVs) in patients at high risk for complications related to reoperation. Patients who undergo VIV TAVR are at risk of patient-prosthesis mismatch, as the transcatheter heart valve (THV) is implanted within the ring of the existing BPV, limiting full expansion and reducing the maximum achievable effective orifice area of the THV. Importantly, patient-prosthesis mismatch and high residual transvalvular gradients are associated with reduced survival following VIV TAVR. Bioprosthetic valve fracture (BVF) is as a novel technique to address this problem. During BPV, a non-compliant valvuloplasty balloon is positioned within the BPV frame, and a highpressure balloon inflation is performed to fracture the surgical sewing ring of the BPV. This allows for further expansion of the BPV as well as the implanted THV, thus increasing the maximum effective orifice area that can be achieved after VIV TAVR. This review focuses on the current evidence base for BVF to facilitate VIV TAVR, including initial bench testing, procedural technique, clinical experience and future directions.

  19. The Chernobyl accident consequences

    International Nuclear Information System (INIS)

    2001-04-01

    Five teen years later, Tchernobyl remains the symbol of the greater industrial nuclear accident. To take stock on this accident, this paper proposes a chronology of the events and presents the opinion of many international and national organizations. It provides also web sites references concerning the environmental and sanitary consequences of the Tchernobyl accident, the economic actions and propositions for the nuclear safety improvement in the East Europe. (A.L.B.)

  20. Radiation, accidents, society

    International Nuclear Information System (INIS)

    1988-01-01

    This book is meant to be used as a reference book for information officers at the event of a nuclear accident. The main part is edited in alphabetical order to facilitate use under stress. The book gives a short review of the health risks of radiation, and descriptions of accidents that have occured. The index words that have been chosen for the main part of the book have been selected due to experiences in connection with incidents and accidents. (L.E.)

  1. Analysis of the primary source term for meltdown accidents using MELCOR 1.8.2

    International Nuclear Information System (INIS)

    Schmuck, P.

    1995-01-01

    The MELCOR code describing accident phenomena in the core and primary systems was used for source term calculations and - in the context of the MELCOR Cooperative Assessment Programme - for studying two-phase flows through components such as valves and chokes. Results of the latter studies in comparison to experiments gave hints for an improved calculation of momentum transfer between the phases. (orig.)

  2. Radio and line transmission 2

    CERN Document Server

    Roddy, Dermot

    2013-01-01

    Radio and Line Transmission, Volume 2 gives a detailed treatment of the subject as well as an introduction to additional advanced subject matter. Organized into 14 chapters, this book begins by explaining the radio wave propagation, signal frequencies, and bandwidth. Subsequent chapters describe the transmission lines and cables; the aerials; tuned and coupled circuits; bipolar transistor amplifiers; field-effect transistors and circuits; thermionic valve amplifiers; LC oscillators; the diode detectors and modulators; and the superheterodyne receiver. Other chapters explore noise and interfere

  3. LOFT pressurizer safety: relief valve reliability

    International Nuclear Information System (INIS)

    Brown, E.S.

    1978-01-01

    The LOFT pressurizer self-actuating safety-relief valves are constructed to the present state-of-the-art and should have reliability equivalent to the valves in use on PWR plants in the U.S. There have been no NRC incident reports on valve failures to lift that would challenge the Technical Specification Safety Limit. Fourteen valves have been reported as lifting a few percentage points outside the +-1% Tech. Spec. surveillance tolerance (9 valves tested over and 5 valves tested under specification). There have been no incident reports on failures to reseat. The LOFT surveillance program for assuring reliability is equivalent to nuclear industry practice

  4. Valve system incorporating single failure protection logic

    Science.gov (United States)

    Ryan, Rodger; Timmerman, Walter J. H.

    1980-01-01

    A valve system incorporating single failure protective logic. The system consists of a valve combination or composite valve which allows actuation or de-actuation of a device such as a hydraulic cylinder or other mechanism, integral with or separate from the valve assembly, by means of three independent input signals combined in a function commonly known as two-out-of-three logic. Using the input signals as independent and redundant actuation/de-actuation signals, a single signal failure, or failure of the corresponding valve or valve set, will neither prevent the desired action, nor cause the undesired action of the mechanism.

  5. LOFT pressurizer safety: relief valve reliability

    Energy Technology Data Exchange (ETDEWEB)

    Brown, E.S.

    1978-01-18

    The LOFT pressurizer self-actuating safety-relief valves are constructed to the present state-of-the-art and should have reliability equivalent to the valves in use on PWR plants in the U.S. There have been no NRC incident reports on valve failures to lift that would challenge the Technical Specification Safety Limit. Fourteen valves have been reported as lifting a few percentage points outside the +-1% Tech. Spec. surveillance tolerance (9 valves tested over and 5 valves tested under specification). There have been no incident reports on failures to reseat. The LOFT surveillance program for assuring reliability is equivalent to nuclear industry practice.

  6. GAMMA-FR and MELCOR Validation using HCS Heat Exchanger Break Accident

    International Nuclear Information System (INIS)

    Jin, Hyung Gon; Hong, Yun Jeong; Cho, Seung Yon

    2016-01-01

    To confirm the HCCR-TBS integrity, enveloped cases from the conceivable events were evaluated and demonstrated compliance with the General Safety Objectives of ITER. In this analysis, amount of discharged helium is the key parameter to examine total tritium ingress to CCWS-1. In this regard, radiation heat transfer and temperature distribution along the pipes did not take account. Due to the same reason, flow network inside of TBM is simplified as one fluid volume (FB1300). In principle, transient of this accident is similar to LOHSA, therefore, TBM temperature is expected to be cool down by passive cooling and isolation valves avoid CCWS-1 pressure build-up during the accident. With relief valve, pressure of CCWS-1 is under 0.43 MPa during LOCA happens. (CCWS-1 max. design pressure: 1MPa). On the other hand, primary concern is tritium concentration increase in CCWS-1 because of tritium contents in HCS coolant. The important point is that CCWS-1 is an ESP device and its ESP level should be confirmed when operating with HCCR-TBS as well. Key parameters, which govern this transient, are relief valve operation, nitrogen in the pressurizer and flow area of the ruptured channels. Relief valve in CCWS-1 pressurizer opens at 0.41 MPa and closes 0.39 MPa, therefore, CCWS-1 pressure is impossible to exceed 0.41 MPa globally. As a comparison, calculation was conducted against CCWS-1 with relief valve (with RV) and without relief valve (without RV)

  7. Reactivity insertion accident analysis

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

    1990-04-01

    The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

  8. Nuclear accidents and epidemiology

    International Nuclear Information System (INIS)

    1987-01-01

    A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

  9. Accidents (FARS) (National)

    Data.gov (United States)

    Department of Transportation — Accident - (1975-current): This data file (NTAD) contains information about crash characteristics and environmental conditions at the time of the crash. There is one...

  10. W-12 valve pit decontamination demonstration

    International Nuclear Information System (INIS)

    Benson, C.E.; Parfitt, J.E.; Patton, B.D.

    1995-12-01

    Waste tank W-12 is a tank in the ORNL Low-Level Liquid Waste (LLLW) system that collected waste from Building 3525. Because of a leaking flange in the discharge line from W-12 to the evaporator service tank (W-22) and continual inleakage into the tank from an unknown source, W-12 was removed from service to comply with the Federal Facilities Agreement requirement. The initial response was to decontaminate the valve pit between tank W-12 and the evaporator service tank (W-22) to determine if personnel could enter the pit to attempt repair of the leaking flange. Preventing the spread of radioactive contamination from the pit to the environment and to other waste systems was of concern during the decontamination. The drain in the pit goes to the process waste system; therefore, if high-level liquid waste were generated during decontamination activities, it would have to be removed from the pit by means other than the available liquid waste connection. Remote decontamination of W-12 was conducted using the General Mills manipulator bridge and telescoping trolley and REMOTEC RM-10 manipulator. The initial objective of repairing the leaking flange was not conducted because of the repair uncertainty and the unknown tank inleakage. Rather, new piping was installed to empty the W-12 tank that would bypass the valve pit and eliminate the need to repair the flange. The radiological surveys indicated that a substantial decontamination factor was achieved

  11. Door valve for fuel handling path

    International Nuclear Information System (INIS)

    Makishima, Katsuhiko.

    1969-01-01

    A door valve is provided which seals cover gas from a liquid metal cooled reactor without leakage therefrom. A threaded shaft is screwed into a heavy box press which is packed with lead. The shaft is adapted to be rotated by an electric motor or a manually operated wheel which is disposed outside of the door valve. A valve plate is suspended from the box press by four guide wheels mounted thereon. The guide wheels are fitted into inclined guide grooves formed at the valve plate and into grooved formed in the inner wall of a valve casing. A locking ball is provided at each side of the valve plate. In operation the shaft rotates and travels to permit the box press and the valve plate to move into the door valve casing, thus releasing the locking balls. The valve plate does not contact the bottom of the casing. When the box press reaches the home position, the valve plate is carried on the valve opening, and the box press presses the valve plate to increase the tightness. The valve plate does not suffer wear as it does not slide over other parts. (Yamaguchi, T.)

  12. Sequential transcatheter aortic valve implantation due to valve dislodgement - a Portico valve implanted over a CoreValve bioprosthesis.

    Science.gov (United States)

    Campante Teles, Rui; Costa, Cátia; Almeida, Manuel; Brito, João; Sondergaard, Lars; Neves, José P; Abecasis, João; M Gabriel, Henrique

    2017-03-01

    Transcatheter aortic valve implantation (TAVI) has become an important treatment in high surgical risk patients with severe aortic stenosis (AS), whose complications need to be managed promptly. The authors report the case of an 86-year-old woman presenting with severe symptomatic AS, rejected for surgery due to advanced age and comorbidities. The patient underwent a first TAVI, with implantation of a Medtronic CoreValve ® , which became dislodged and migrated to the ascending aorta. Due to the previous balloon valvuloplasty, the patient's AS became moderate, and her symptoms improved. After several months, she required another intervention, performed with a St. Jude Portico ® repositionable self-expanding transcatheter aortic valve. There was a good clinical response that was maintained at one-year follow-up. The use of a self-expanding transcatheter bioprosthesis with repositioning features is a solution in cases of valve dislocation to avoid suboptimal positioning of a second implant, especially when the two valves have to be positioned overlapping or partially overlapping each other. Copyright © 2017 Sociedade Portuguesa de Cardiologia. Publicado por Elsevier España, S.L.U. All rights reserved.

  13. Cavitation noise from butterfly valves

    International Nuclear Information System (INIS)

    Rahmeyer, W.J.

    1982-01-01

    Cavitation in valves can produce levels of intense noise. It is possible to mathematically express a limit for a design level of cavitation noise in terms of the cavitation parameter sigma. Using the cavitation parameter or limit, it is then possible to calculate the flow conditions at which a design level of cavitation noise will occur. However, the intensity of cavitation increases with the upstream pressure and valve size at a constant sigma. Therefore, it is necessary to derive equations to correct or scale the cavitation limit for the effects of different upstream pressures and valve sizes. The following paper discusses and presents experimental data for the caviation noise limit as well as the cavitation limits of incipient, critical, incipient damage, and choking cavitation for butterfly valves. The main emphasis is on the design limit of caviation noise, and a noise level of 85 decibels was selected as the noise limit. Tables of data and scaling exponents are included for applying the design limits for the effects of upstream pressure and valve size. (orig.)

  14. Plunger with simple retention valve

    International Nuclear Information System (INIS)

    Fekete, A.V.

    1987-01-01

    This patent describes a positive displacement retention valve apparatus in which the actual flow equals the theoretical maximum flow through the retention valve. The apparatus includes, in combination, a confined fluid flow conduit, a piston adapted for reciprocal movement within the fluid flow conduit between upstream and downstream limit positions, piston reciprocating means, and pressure responsive check valve means located upstream with respect to the piston in the fluid flow conduit. The pressure responsive check valve means operable to permit fluid flow therethrough in a downstream direction toward the piston, and to preclude fluid flow therethrough in an opposite direction. The piston is composed of parts which are relatively movable with respect to one another. The piston includes a simple retention valve consisting of a plug means, a cylinder having a minimum and a maximum internal cross section flow area therein and being reciprocal within the confined fluid flow conduit, and a seat on the cylinder for the plug means. The piston reciprocating means are operatively connected to the plug means

  15. Structural valve deterioration in the Mitroflow biological heart valve prosthesis

    DEFF Research Database (Denmark)

    Issa, Issa Farah; Poulsen, Steen Hvitfeldt; Waziri, Farhad

    2018-01-01

    OBJECTIVES: Concern has been raised regarding the long-term durability of the Mitroflow biological heart valve prosthesis. Our aim was to assess the incidence of structural valve degeneration (SVD) for the Mitroflow bioprosthesis in a nationwide study in Denmark including all patients alive......: A total of 173 patients were diagnosed with SVD by echocardiography. Of these, 64 (11%) patients had severe SVD and 109 (19%) patients moderate SVD. Severe SVD was associated with the age of the prosthesis and small prosthesis size [Size 21: hazard ratio (95% confidence interval, CI) 2.72 (0.97-8.56), P...

  16. Passive depressurization accident management strategy for boiling water reactors

    International Nuclear Information System (INIS)

    Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

    2015-01-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

  17. Passive depressurization accident management strategy for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

    2015-04-01

    Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

  18. Fibrin glue on an aortic cusp detected by transesophageal echocardiography after valve-sparing aortic valve replacement: a case report.

    Science.gov (United States)

    Nakahira, Junko; Ishii, Hisanari; Sawai, Toshiyuki; Minami, Toshiaki

    2015-03-07

    Fibrin glue is used commonly during cardiac surgery but can behave as an intracardiac abnormal foreign body following surgery. There have been few such cases reported, and they were typically noticed only because of the resulting catastrophic cardiac conditions, such as valvular malfunction. We report a case where, for the first time, transesophageal echocardiography was used to detected fibrin glue that was adherent to the ventricular side of a patient's aortic valve immediately after aortic declamping. A 45-year-old Japanese man with Marfan syndrome underwent an aortic valve-sparing operation to treat moderate aortic valve regurgitation resulting from enlargement of his right coronary cusp. Fibrin glue was lightly applied to the suture line between the previous and new grafts. Transesophageal echocardiography performed prior to weaning from the cardiopulmonary bypass revealed mild aortic valve regurgitation in addition to a mobile membranous structure attached to the ventricular side of his aortic valve. It was identified as fibrin glue. We resolved the regurgitation by removing the fibrin glue and repeating the aortic cusp plication. The patient had no complications during recovery. Fibrin glue can act as an intracardiac foreign body and lead to a potentially fatal embolism. We demonstrated the use of transesophageal echocardiography to detect a fibrin glue-derived intracardiac abnormal foreign body and to confirm its removal. To the best of our knowledge, this is the first case where fibrin glue adherent to the aortic valve was detected by transesophageal echocardiography. These findings demonstrate the importance of using transesophageal echocardiography during cardiac surgery that involves using biological glues.

  19. Effectiveness of electronic stability control on single-vehicle accidents

    DEFF Research Database (Denmark)

    Lyckegaard, Allan; Hels, Tove; Bernhoft, Inger Marie

    2015-01-01

    the injury severity categories (slight, severe, and fatal). Conclusions: In line with previous results, this study concludes that ESC reduces the risk for single-vehicle injury accidents by 31% when controlling for various confounding factors related to the driver, the car, and the accident surroundings......Objective: This study aims at evaluating the effectiveness of electronic stability control (ESC) on single-vehicle injury accidents while controlling for a number of confounders influencing the accident risk. Methods: Using police-registered injury accidents from 2004 to 2011 in Denmark with cars...... the following were significant. For the driver: Age, gender, driving experience, valid driving license, and seat belt use. For the vehicle: Year of registration, weight, and ESC. For the accident surroundings: Visibility, light, and location. Finally, for the road: Speed limit, surface, and section...

  20. Domino effect in chemical accidents: main features and accident sequences

    OpenAIRE

    Casal Fàbrega, Joaquim; Darbra Roman, Rosa Maria

    2010-01-01

    The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes a...

  1. Qualification by analogy of the functional valving of French pressurized water nuclear power stations

    International Nuclear Information System (INIS)

    Grenet, M.

    1991-01-01

    In certain postulated accidental conditions (loss of coolant accident or secondary pipe rupture, earthquake, high energy pipe rupture) plant valving is called on the important functions to bring the reactor to and maintain it at a safe shutdown condition. ELWCTRICITE DE FRANCE has completed qualification tests of about forty valves to assure their operability. However, taking into account the costs and time required to obtain this qualification and the number of valves to be qualified, this method alone is not sufficient. For this reason, Electricite de France has developed the alternative qualification methodology by analogy for each postulated accidental situation. Feedback experience of these methods today is such that it can be they have achieved their objective; namely, to improve the safety of French pressurized water nuclear power stations, while at the same time avoiding the two dangers represented by excessive complexity resulting in unsatisfactory operation, and insufficient thoroughness not providing any real increase in safety. (author)

  2. Load-dependent extracellular matrix organization in atrioventricular heart valves: differences and similarities.

    Science.gov (United States)

    Alavi, S Hamed; Sinha, Aditi; Steward, Earl; Milliken, Jeffrey C; Kheradvar, Arash

    2015-07-15

    The extracellular matrix of the atrioventricular (AV) valves' leaflets has a key role in the ability of these valves to properly remodel in response to constantly varying physiological loads. While the loading on mitral and tricuspid valves is significantly different, no information is available on how collagen fibers change their orientation in response to these loads. This study delineates the effect of physiological loading on AV valves' leaflets microstructures using Second Harmonic Generation (SHG) microscopy. Fresh natural porcine tricuspid and mitral valves' leaflets (n = 12/valve type) were cut and prepared for the experiments. Histology and immunohistochemistry were performed to compare the microstructural differences between the valves. The specimens were imaged live during the relaxed, loading, and unloading phases using SHG microscopy. The images were analyzed with Fourier decomposition to mathematically seek changes in collagen fiber orientation. Despite the similarities in both AV valves as seen in the histology and immunohistochemistry data, the microstructural arrangement, especially the collagen fiber distribution and orientation in the stress-free condition, were found to be different. Uniaxial loading was dependent on the arrangement of the fibers in their relaxed mode, which led the fibers to reorient in-line with the load throughout the depth of the mitral leaflet but only to reorient in-line with the load in deeper layers of the tricuspid leaflet. Biaxial loading arranged the fibers in between the two principal axes of the stresses independently from their relaxed states. Unlike previous findings, this study concludes that the AV valves' three-dimensional extracellular fiber arrangement is significantly different in their stress-free and uniaxially loaded states; however, fiber rearrangement in response to the biaxial loading remains similar. Copyright © 2015 the American Physiological Society.

  3. Bicuspid Aortic Valve Disease: A Comprehensive Review

    OpenAIRE

    Mordi, Ify; Tzemos, Nikolaos

    2012-01-01

    Bicuspid aortic valve is the commonest congenital cardiac abnormality in the general population. This paper article will discuss our current knowledge of the anatomy, pathophysiology, genetics, and clinical aspects of bicuspid aortic valve disease.

  4. Echocardiographic evaluation of heart valve prosthetic dysfunction

    Directory of Open Access Journals (Sweden)

    Yuriy Ivaniv

    2018-02-01

    Full Text Available Patients with replaced heart valve submitted to echocardiographic examination may have symptoms related either to valvular malfunction or ventricular dysfunction from different causes. Clinical examination is not reliable in a prosthetic valve evaluation and the main information regarding its function could be obtained using different cardiac ultrasound modalities. This review provides a description of echocardiographic and Doppler techniques useful in evaluation of prosthetic heart valves. For the interpretation of echocardiography there is a need in special knowledge of prosthesis types and possible reasons of prosthetic function deterioration. Echocardiography allows to reveal valve thrombosis, pannus formation, vegetation and such complications of infective endocarditis as valve ring abscess or dehiscence. Transthoracic echocardiography requires different section plane angles and unconventional views. Transesophageal echocardiography is more often used than in native valve examination due to better visualization of prosthetic valve structure and function. Three-dimensional echocardiography could provide more detailed visual information especially in the assessment of paravalvular regurgitation or valve obstruction.

  5. Bistable fluidic valve is electrically switched

    Science.gov (United States)

    Fiet, O.; Salvinski, R. J.

    1970-01-01

    Bistable control valve is selectively switched by direct application of an electrical field to divert fluid from one output channel to another. Valve is inexpensive, has no moving parts, and operates on fluids which are relatively poor electrical conductors.

  6. Comparative study between CardiaMed valves (freely floating valve leaflets versus St. Jude Medical (fixed valve leaflets in mitral valve replacement surgery

    Directory of Open Access Journals (Sweden)

    Mostafa Ahmed

    2017-09-01

    Conclusions: CardiaMed freely floating leaflet prostheses showed good hemodynamic characteristics. The prosthesis adequately corrects hemodynamics and is safe and no worse than the St. Jude Medical valve in the mitral valve position.

  7. Prosthetic Mitral Valve Leaflet Escape

    Science.gov (United States)

    Kim, Darae; Hun, Sin Sang; Cho, In-Jeong; Shim, Chi-Young; Ha, Jong-Won; Chung, Namsik; Ju, Hyun Chul; Sohn, Jang Won

    2013-01-01

    Leaflet escape of prosthetic valve is rare but potentially life threatening. It is essential to make timely diagnosis in order to avoid mortality. Transesophageal echocardiography and cinefluoroscopy is usually diagnostic and the location of the missing leaflet can be identified by computed tomography (CT). Emergent surgical correction is mandatory. We report a case of fractured escape of Edward-Duromedics mitral valve 27 years after the surgery. The patient presented with symptoms of acute decompensated heart failure and cardiogenic shock. She was instantly intubated and mechanically ventilated. After prompt evaluation including transthoracic echocardiography and CT, the escape of the leaflet was confirmed. The patient underwent emergent surgery for replacement of the damaged prosthetic valves immediately. Eleven days after the surgery, the dislodged leaflet in iliac artery was removed safely and the patient recovered well. PMID:23837121

  8. Active combustion flow modulation valve

    Science.gov (United States)

    Hensel, John Peter; Black, Nathaniel; Thorton, Jimmy Dean; Vipperman, Jeffrey Stuart; Lambeth, David N; Clark, William W

    2013-09-24

    A flow modulation valve has a slidably translating hollow armature with at least one energizable coil wound around and fixably attached to the hollow armature. The energizable coil or coils are influenced by at least one permanent magnet surrounding the hollow armature and supported by an outer casing. Lorentz forces on the energizable coils which are translated to the hollow armature, increase or decrease the flow area to provide flow throttling action. The extent of hollow armature translation depends on the value of current supplied and the direction of translation depends on the direction of current flow. The compact nature of the flow modulation valve combined with the high forces afforded by the actuator design provide a flow modulation valve which is highly responsive to high-rate input control signals.

  9. Statins for aortic valve stenosis

    Directory of Open Access Journals (Sweden)

    Luciana Thiago

    Full Text Available ABSTRACT BACKGROUND: Aortic valve stenosis is the most common type of valvular heart disease in the USA and Europe. Aortic valve stenosis is considered similar to atherosclerotic disease. Some studies have evaluated statins for aortic valve stenosis. OBJECTIVES: To evaluate the effectiveness and safety of statins in aortic valve stenosis. METHODS: Search methods: We searched the Cochrane Central Register of Controlled Trials (CENTRAL, MEDLINE, Embase, LILACS - IBECS, Web of Science and CINAHL Plus. These databases were searched from their inception to 24 November 2015. We also searched trials in registers for ongoing trials. We used no language restrictions. Selection criteria: Randomized controlled clinical trials (RCTs comparing statins alone or in association with other systemic drugs to reduce cholesterol levels versus placebo or usual care. Data collection and analysis: Primary outcomes were severity of aortic valve stenosis (evaluated by echocardiographic criteria: mean pressure gradient, valve area and aortic jet velocity, freedom from valve replacement and death from cardiovascular cause. Secondary outcomes were hospitalization for any reason, overall mortality, adverse events and patient quality of life. Two review authors independently selected trials for inclusion, extracted data and assessed the risk of bias. The GRADE methodology was employed to assess the quality of result findings and the GRADE profiler (GRADEPRO was used to import data from Review Manager 5.3 to create a 'Summary of findings' table. MAIN RESULTS: We included four RCTs with 2360 participants comparing statins (1185 participants with placebo (1175 participants. We found low-quality evidence for our primary outcome of severity of aortic valve stenosis, evaluated by mean pressure gradient (mean difference (MD -0.54, 95% confidence interval (CI -1.88 to 0.80; participants = 1935; studies = 2, valve area (MD -0.07, 95% CI -0.28 to 0.14; participants = 127; studies = 2

  10. Statins for aortic valve stenosis.

    Science.gov (United States)

    Thiago, Luciana; Tsuji, Selma Rumiko; Nyong, Jonathan; Puga, Maria Eduarda Dos Santos; Góis, Aécio Flávio Teixeira de; Macedo, Cristiane Rufino; Valente, Orsine; Atallah, Álvaro Nagib

    2016-01-01

    Aortic valve stenosis is the most common type of valvular heart disease in the USA and Europe. Aortic valve stenosis is considered similar to atherosclerotic disease. Some studies have evaluated statins for aortic valve stenosis. To evaluate the effectiveness and safety of statins in aortic valve stenosis. Search methods: We searched the Cochrane Central Register of Controlled Trials (CENTRAL), MEDLINE, Embase, LILACS - IBECS, Web of Science and CINAHL Plus. These databases were searched from their inception to 24 November 2015. We also searched trials in registers for ongoing trials. We used no language restrictions.Selection criteria: Randomized controlled clinical trials (RCTs) comparing statins alone or in association with other systemic drugs to reduce cholesterol levels versus placebo or usual care. Data collection and analysis: Primary outcomes were severity of aortic valve stenosis (evaluated by echocardiographic criteria: mean pressure gradient, valve area and aortic jet velocity), freedom from valve replacement and death from cardiovascular cause. Secondary outcomes were hospitalization for any reason, overall mortality, adverse events and patient quality of life.Two review authors independently selected trials for inclusion, extracted data and assessed the risk of bias. The GRADE methodology was employed to assess the quality of result findings and the GRADE profiler (GRADEPRO) was used to import data from Review Manager 5.3 to create a 'Summary of findings' table. We included four RCTs with 2360 participants comparing statins (1185 participants) with placebo (1175 participants). We found low-quality evidence for our primary outcome of severity of aortic valve stenosis, evaluated by mean pressure gradient (mean difference (MD) -0.54, 95% confidence interval (CI) -1.88 to 0.80; participants = 1935; studies = 2), valve area (MD -0.07, 95% CI -0.28 to 0.14; participants = 127; studies = 2), and aortic jet velocity (MD -0.06, 95% CI -0.26 to 0

  11. Evaluation of mispositioned ECCS valves

    International Nuclear Information System (INIS)

    Hill, R.A.; O'Brien, J.F.; McIntire, D.C.; Barlow, R.T.

    1977-09-01

    In October of 1975, Westinghouse submitted NS-CE-787, dated October 17, 1975, to the Nuclear Regulatory Commission (NRC) and entered into discussions with them concerning the spurious movement of certain motor-operated valves (MOV's) in the Emergency Core Cooling System (ECCS) to a position defeating the ECCS function at a time when this function is required. On November 25, 1975, the discussion turned to the possible movement of a manually controlled, motor-operated valve due to a fault in its electrical circuitry and the NRC staff expressed concerns about other possible failure modes that might lead to such a valve movement. The NRC meeting minutes document these concerns. This report is an item-by-item response to the concerns expressed by the NRC staff at that meeting and incorporates the original electrical fault analysis

  12. Novel Active Combustion Control Valve

    Science.gov (United States)

    Caspermeyer, Matt

    2014-01-01

    This project presents an innovative solution for active combustion control. Relative to the state of the art, this concept provides frequency modulation (greater than 1,000 Hz) in combination with high-amplitude modulation (in excess of 30 percent flow) and can be adapted to a large range of fuel injector sizes. Existing valves often have low flow modulation strength. To achieve higher flow modulation requires excessively large valves or too much electrical power to be practical. This active combustion control valve (ACCV) has high-frequency and -amplitude modulation, consumes low electrical power, is closely coupled with the fuel injector for modulation strength, and is practical in size and weight. By mitigating combustion instabilities at higher frequencies than have been previously achieved (approximately 1,000 Hz), this new technology enables gas turbines to run at operating points that produce lower emissions and higher performance.

  13. Pannus Formation Leads to Valve Malfunction in the Tricuspid Position 19 Years after Triple Valve Replacement.

    Science.gov (United States)

    Alskaf, Ebraham; McConkey, Hannah; Laskar, Nabila; Kardos, Attila

    2016-06-20

    The Medtronic ATS Open Pivot mechanical valve has been successfully used in heart valve surgery for more than two decades. We present the case of a patient who, 19 years following a tricuspid valve replacement with an ATS prosthesis as part of a triple valve operation following infective endocarditis, developed severe tricuspid regurgitation due to pannus formation.

  14. The nordic aortic valve intervention (NOTION) trial comparing transcatheter versus surgical valve implantation

    DEFF Research Database (Denmark)

    Thyregod, Hans Gustav; Søndergaard, Lars; Ihlemann, Nikolaj

    2013-01-01

    Degenerative aortic valve (AV) stenosis is the most prevalent heart valve disease in the western world. Surgical aortic valve replacement (SAVR) has until recently been the standard of treatment for patients with severe AV stenosis. Whether transcatheter aortic valve implantation (TAVI) can...

  15. Early clinical outcome of aortic transcatheter valve-in-valve implantation in the Nordic countries

    DEFF Research Database (Denmark)

    Ihlberg, Leo; Nissen, Henrik Hoffmann; Nielsen, Niels Erik

    2013-01-01

    Transcatheter valve-in-valve implantation has emerged as an option, in addition to reoperative surgical aortic valve replacement, to treat failed biologic heart valve substitutes. However, the clinical experience with this approach is still limited. We report the comprehensive experience...

  16. Intro to Valve Guide Reconditioning. Automotive Mechanics. Valves. Instructor's Guide [and] Student Guide.

    Science.gov (United States)

    Horner, W.

    This instructional package, one in a series of individualized instructional units on tools and techniques for repairing worn valve guides in motor vehicles, provides practical experience for students in working on cylinder heads. Covered in the module are reaming valve guides that are oversized to match a new oversized valve, reaming valve guides…

  17. Infective Endocarditis of the Aortic Valve with Anterior Mitral Valve Leaflet Aneurysm

    NARCIS (Netherlands)

    Tomsic, Anton; Li, Wilson W. L.; van Paridon, Marieke; Bindraban, Navin R.; de Mol, Bas A. J. M.

    2016-01-01

    Mitral valve leaflet aneurysm is a rare and potentially devastating complication of aortic valve endocarditis. We report the case of a 48-year-old man who had endocarditis of the native aortic valve and a concomitant aneurysm of the anterior mitral valve leaflet. Severe mitral regurgitation occurred

  18. Chernobyl accident and Denmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. Volume 2 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  19. Criticality accident in Argentina

    International Nuclear Information System (INIS)

    Oliveira, A.R. de.

    1984-01-01

    A recent criticality type accident, ocurred in Argetina, is commented. Considerations about the nature of the facility where this accident took place, its genesis, type of operation carried out on the day of the event, and the medical aspects involved are done. (Author) [pt

  20. Chernobyl accident and Danmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by the Secretary of State for the Environment. Volume 1 contains copies of original documents issued by Danish authorities during the first accident phase and afterwards. Evaluations, monitoring data, press releases, legislation acts etc. are included. (author)

  1. Communication and industrial accidents

    NARCIS (Netherlands)

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational communication on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. As a link between these two levels - the organizational failures and mistakes - I suggest the

  2. Chapter 6: Accidents

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2018-04-01

    Th chapter 6 presents the accidents of: 1) Stimos (Italy - May, 1975); 2) San Salvador (El Salvador - February 5, 1989); 3) Soreq (Israel - June 21, 1990); 4) Nesvizh (Belarus - October 26, 1991); 5) Illinois (USA - February, 1965); 6)Maryland (EUA - December 11, 1991); 7)Hanoi (Vietnam -November 17, 1992); 8)Fleurus (Belgium - March 11, 2006) and final remarks on accidents.

  3. Thermalydraulic processes in the reactor coolant system of a BWR under severe accident conditions

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1990-01-01

    Boiling water reactors (BWRs) incorporate many unique structural features that make their expected response under severe accident conditions very different from that predicted in the case of pressurized water reactor accident sequences. Automatic main steam isolation valve (MIV) closure as the vessel water level approaches the top of the core would cause reactor vessel isolation while automatic recirculation pump trip would limit the in-vessel flows to those characteristic of natural circulation (as disturbed by vessel relief valve actuation). This paper provides a discussion of the BWR control blade, channel box, core plate, control rod guide tube, and reactor vessel safety relief valve (SRV) configuration and the effects of these structural components upon thermal hydraulic processes within the reactor vessel under severe accident conditions. The dominant BWR severe accident sequences as determined by probabilistic risk assessment are described and the expected timing of events for the unmitigated short-term station blackout severe accident sequence at the Peach Bottom atomic power station is presented

  4. Radiological accidents in medical practice

    International Nuclear Information System (INIS)

    Cardenas Herrera, Juan

    2012-01-01

    Different radiological accidents that may occur in medical practice are shown. The following topics are focused: accident statistics for medical exposure, accidental medical exposures, radiotherapy accidents and potential accidental scenarios [es

  5. Dangers of bypassing thermal overload relays in nuclear power plant motor operated valve circuits

    International Nuclear Information System (INIS)

    Baxter, F.D.

    1980-01-01

    Operation of motor operated valves is analyzed under various abnormal conditions such as frozen bearing, tight packing, mid-travel obstruction, torque switch failure, limit switch failure, and post-accident operation. Each condition has been reviewed to show that an adverse situation results if the thermal overload relays in the circuit are bypassed. In conclusion, there appears to be no technical basis for bypassing or oversizing the thermal overload relay provided it is selected correctly

  6. Check valve slam waterhammer in piping systems equipped with multiple parallel pumps

    International Nuclear Information System (INIS)

    Sponsel, J.; Bird, E.; Zarechnak, A.

    1993-01-01

    The low pressure safety injection system at the calvert cliff's plant is designed to provide cooling water to the reactor in the event of a postulated accident and for reactor cool-down and decay heat removal during normal maintenance and refueling. This system experienced repeated damage to the axial piping supports on the pump section and the discharge headers due to the check valve phenomenon. To determine the cause, testing was performed in both the LPSI and CCW systems

  7. Small sodium valve design and operating experience

    International Nuclear Information System (INIS)

    McGough, C.B.

    1974-01-01

    The United States Liquid Metal Fast Breeder Reactor program (LMFBR) includes an extensive program devoted to the development of small sodium valves. This program is now focused on the development and production of valves for the Fast Flux Test Facility (FFTF) now under construction near Richland, Washington. Other AEC support facilities, such as various test loops located at the Liquid Metal Engineering Center (LMEC), Los Angeles, California, and at the Hanford Engineering Development Laboratory (HEDL), Richland, Washington, also have significant requirements for small sodium valves, and valves similar in design to the FFTF valves are being supplied to these AEC laboratories for use in their critical test installations. A principal motivation for these valve programs, beyond the immediate need to provide high-reliability valves for FFTF and the support facilities, is the necessity to develop small valve technology for the Clinch River Breeder Reactor Plant (CRBRP). FFTF small sodium valve design and development experience will be directly applied to the CRBRP program. Various test programs have been, and are being, conducted to verify the performance and integrity of the FFTF valves, and to uncover any potential problems so that they can be corrected before the valves are placed in service in FFTF. The principal small sodium valve designs being utilized in current U.S. programs, the test and operational experience obtained to date on them, problems uncovered, and future development and testing efforts being planned are reviewed. The standards and requirements to which the valves are being designed and fabricated, the valve designs in current use, valve operators, test and operating experience, and future valve development plans are summarized. (U.S.)

  8. Promising results after percutaneous mitral valve repair

    DEFF Research Database (Denmark)

    Ihlemann, Nikolaj; Franzen, Olaf; Jørgensen, Erik

    2011-01-01

    Mitral valve regurgitation (MR) is the secondmost frequent valve disease in Europe. Untreated MR causes considerable morbidity and mortality. In the elderly, as many as half of these patients are denied surgery because of an estimated high surgical risk. Percutaneous mitral valve repair with the ...... with the MitraClip system resembles the Alfieristitch where a clip is used to connect the tip of the mitral valve leaflets....

  9. Infective endocarditis following percutaneous pulmonary valve replacement

    DEFF Research Database (Denmark)

    Cheung, Gary; Vejlstrup, Niels; Ihlemann, Nikolaj

    2013-01-01

    Infective endocarditis (IE) following percutaneous pulmonary valve replacement (PPVR) with the Melody valve is rarely reported. Furthermore, there are challenges in this diagnosis; especially echocardiographic evidence of vegetation within the prosthesis may be difficult.......Infective endocarditis (IE) following percutaneous pulmonary valve replacement (PPVR) with the Melody valve is rarely reported. Furthermore, there are challenges in this diagnosis; especially echocardiographic evidence of vegetation within the prosthesis may be difficult....

  10. Fast-acting valve actuator

    Science.gov (United States)

    Cho, Nakwon

    1980-01-01

    A fast-acting valve actuator utilizes a spring driven pneumatically loaded piston to drive a valve gate. Rapid exhaust of pressurized gas from the pneumatically loaded side of the piston facilitates an extremely rapid piston stroke. A flexible selector diaphragm opens and closes an exhaust port in response to pressure differentials created by energizing and de-energizing a solenoid which controls the pneumatic input to the actuator as well as selectively providing a venting action to one side of the selector diaphragm.

  11. Effects of the blockage ratio of a valve disk on loss coefficient in a butterfly valve

    International Nuclear Information System (INIS)

    Rho, Hyung Joon; Lee, Jee Keun; Choi, Hee Joo

    2008-01-01

    The loss coefficient of the butterfly valve which allows partial opening of the valve at closed position and is applicable to the small-sized pipe system with the diameter of 1 inch was measured for the variation of the valve disk blockage ratio. Two different types of the valve disk configuration to adjust the blockage ratio were considered. One was the solid type valve disk of which the diameter was changed into the smaller size rather than the pipe diameter, and the other was the perforate type valve disk on which some holes were perforated. The results from two types of valve disk were compared to identify their characteristics in the loss coefficient distributions. The loss coefficient and the controllable angle of the valve disk were decreased exponentially with the decrease of the blockage ratio. In addition, the perforate valve disk had the effect on the higher loss coefficient rather than the solid type valve disk

  12. 49 CFR 195.260 - Valves: Location.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Valves: Location. 195.260 Section 195.260... PIPELINE Construction § 195.260 Valves: Location. A valve must be installed at each of the following locations: (a) On the suction end and the discharge end of a pump station in a manner that permits isolation...

  13. Porcine Tricuspid Valve Anatomy and Human Compatibility

    DEFF Research Database (Denmark)

    Waziri, Farhad; Lyager Nielsen, Sten; Hasenkam, J. Michael

    2016-01-01

    before clinical use. The study aim was to evaluate and compare the tricuspid valve anatomy of porcine and human hearts. METHODS: The anatomy of the tricuspid valve and the surrounding structures that affect the valve during a cardiac cycle were examined in detail in 100 fresh and 19 formalin...

  14. Valve-sparing aortic root replacement†

    NARCIS (Netherlands)

    Koolbergen, David R.; Manshanden, Johan S. J.; Bouma, Berto J.; Blom, Nico A.; Mulder, Barbara J. M.; de Mol, Bas A. J. M.; Hazekamp, Mark G.

    2015-01-01

    To evaluate our results of valve-sparing aortic root replacement and associated (multiple) valve repair. From September 2003 to September 2013, 97 patients had valve-sparing aortic root replacement procedures. Patient records and preoperative, postoperative and recent echocardiograms were reviewed.

  15. Solving the problem of valve stem leakage

    International Nuclear Information System (INIS)

    Dixon, D.F.

    1976-01-01

    Engineering solutions to valve stem leakage, in systems carrying expensive heavy water under pressure, have progressed from changing packing brands (failure) to leak collection (partial success) to elimination of small packed valves and an improved valve packing strategy involving stable packing materials, live Belleville spring-loading of packing, and issuance of a detailed stuffing box specification (success). (E.C.B.)

  16. 49 CFR 229.109 - Safety valves.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Safety valves. 229.109 Section 229.109..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Steam Generators § 229.109 Safety valves. Every steam generator shall be equipped with at least two safety valves that have a...

  17. Miniature piezo electric vacuum inlet valve

    Science.gov (United States)

    Keville, Robert F.; Dietrich, Daniel D.

    1998-03-24

    A miniature piezo electric vacuum inlet valve having a fast pulse rate and is battery operated with variable flow capability. The low power (piezo electric valves which require preloading of the crystal drive mechanism and 120 Vac, thus the valve of the present invention is smaller by a factor of three.

  18. Valve Corporation: Strategy Tipping Points and Thresholds

    OpenAIRE

    Teppo Felin

    2015-01-01

    Valve Corporation represents an intriguing case study of flat structure and self organization (Puranam & Håkonsson, 2015; Valve, 2012).  The structures and practices of Valve of course are not new. But the company provides an interesting experiment and illustration that powerfully highlights how organizational design can impact individual and collective behavior, strategy and performance.

  19. Valve Corporation: Strategy Tipping Points and Thresholds

    Directory of Open Access Journals (Sweden)

    Teppo Felin

    2015-06-01

    Full Text Available Valve Corporation represents an intriguing case study of flat structure and self organization (Puranam & Håkonsson, 2015; Valve, 2012.  The structures and practices of Valve of course are not new. But the company provides an interesting experiment and illustration that powerfully highlights how organizational design can impact individual and collective behavior, strategy and performance.

  20. Door valve for fuel handling path

    International Nuclear Information System (INIS)

    Makishima, Katsuhiko.

    1969-01-01

    A door valve is provided which seals cover gas from a liquid metal cooled reactor without leakage therefrom. A threaded shaft is screwed into a heavy box press which is packed with lead. The shaft is adapted to be rotated by an electric motor or a manually operated wheel which is disposed outside of the door valve. From the box press a valve plate is suspended by four linkage bars, one for each corner. Each linkage bar is provided with two wheels which are respectively mounted at the connections with the box press and the valve plate. The wheels are carried on the horizontal grooves formed in a door valve casing. In operation the shaft rotates and travels to permit the box press and the valve plate to move into the door valve casing while the valve plate does not contact the casing. When the box press reaches the home position, the wheels drop into the recesses which are disposed at the ends of the grooves, the valve plate is carried on the valve opening, and the box press presses the valve plate to increase the tightness. The valve plate does not suffer wear as it does not over other parts. (Yamaguchi, T.)

  1. Sequential transcatheter aortic valve implantation due to valve dislodgement

    DEFF Research Database (Denmark)

    Campante Teles, Rui; Costa, Cátia; Almeida, Manuel

    2017-01-01

    Transcatheter aortic valve implantation (TAVI) has become an important treatment in high surgical risk patients with severe aortic stenosis (AS), whose complications need to be managed promptly. The authors report the case of an 86-year-old woman presenting with severe symptomatic AS, rejected...

  2. [Accidents and injuries at work].

    Science.gov (United States)

    Standke, W

    2014-06-01

    In the case of an accident at work, the person concerned is insured by law according to the guidelines of the Sozialgesetzbuch VII as far as the injuries have been caused by this accident. The most important source of information on the incident in question is the accident report that has to be sent to the responsible institution for statutory accident insurance and prevention by the employer, if the accident of the injured person is fatal or leads to an incapacity to work for more than 3 days (= reportable accident). Data concerning accidents like these are sent to the Deutsche Gesetzliche Unfallversicherung (DGUV) as part of a random sample survey by the institutions for statutory accident insurance and prevention and are analyzed statistically. Thus the key issues of accidents can be established and used for effective prevention. Although the success of effective accident prevention is undisputed, there were still 919,025 occupational accidents in 2011, with clear gender-related differences. Most occupational accidents involve the upper and lower extremities. Accidents are analyzed comprehensively and the results are published and made available to all interested parties in an effort to improve public awareness of possible accidents. Apart from reportable accidents, data on the new occupational accident pensions are also gathered and analyzed statistically. Thus, additional information is gained on accidents with extremely serious consequences and partly permanent injuries for the accident victims.

  3. Regulatory standpoints on the design-basis capability of safety-related motor-operated valves(MOVs) and power-operated gate valves(POGVs)

    International Nuclear Information System (INIS)

    Kim, W. T.; Kum, O. H.

    1999-01-01

    The weakness in the design-basis capability of Motor-Operated Valves(MOVs) and the susceptibility to Pressure Locking and Thermal Binding phenomena of Power-Operated Gate Valves(POGVs) have been major concerns to be resolved in the nuclear society in and abroad since Three Mile Island accident occurred in the USA in 1979. Through detailed analysis of operating experience and regulatory activities, some MOVs and POGVs have been found to be unreliable in performing their safety functions when they are required to do so under certain conditions, especially under design-basis accident conditions. Further, it is well understood that these safety problems may not be identified by the typical valve in-service testing(IST). USNRC has published three Generic Letters, GL 89-10, GL 95-07, and GL 96-05, requiring nuclear plant licensees to take appropriate actions to resolve the problems mentioned above. Korean nuclear regulatory body has made public an administration measure called 'Regulatory recommendation to verify safety functions of the safety-related MOVs and POGVs' on June 13, 1997, and in this administration measure Korean utility is asked to submit written documents to show how it assure design-basis capability of these valves. The following are among the major concerns being considered from a regulation standpoint. Program scope and implementation priority, dynamic tests under differential pressure conditions, accuracy of diagnostic equipment, torque switch setting and torque bypass percentage, weak link analysis, motor actuator sizing, corrective actions taken to resolve pressure locking and thermal binding susceptibility, and a periodic verification program for the valves once design-basis capability has been verified

  4. Supra-annular valve strategy for an early degenerated transcatheter balloon-expandable heart valve.

    Science.gov (United States)

    Kamioka, Norihiko; Caughron, Hope; Corrigan, Frank; Block, Peter; Babaliaros, Vasilis

    2018-01-23

    Currently, there are no recommendations regarding the selection of valve type for a transcatheter heart valve (THV)-in-THV procedure. A supra-annular valve design may be superior in that it results in a larger effective orifice area and may have a lower chance of valve thrombosis after THV-in-THV. In this report, we describe the use of a supra-annular valve strategy for an early degenerated THV. © 2018 Wiley Periodicals, Inc.

  5. Aortic valve insufficiency in the teenager and young adult: the role of prosthetic valve replacement.

    Science.gov (United States)

    Bradley, Scott M

    2013-10-01

    The contents of this article were presented in the session "Aortic insufficiency in the teenager" at the congenital parallel symposium of the 2013 Society of Thoracic Surgeons (STS) annual meeting. The accompanying articles detail the approaches of aortic valve repair and the Ross procedure.(1,2) The current article focuses on prosthetic valve replacement. For many young patients requiring aortic valve surgery, either aortic valve repair or a Ross procedure provides a good option. The advantages include avoidance of anticoagulation and potential for growth. In other patients, a prosthetic valve is an appropriate alternative. This article discusses the current state of knowledge regarding mechanical and bioprosthetic valve prostheses and their specific advantages relative to valve repair or a Ross procedure. In current practice, young patients requiring aortic valve surgery frequently undergo valve replacement with a prosthetic valve. In STS adult cardiac database, among patients ≤30 years of age undergoing aortic valve surgery, 34% had placement of a mechanical valve, 51% had placement of a bioprosthetic valve, 9% had aortic valve repair, and 2% had a Ross procedure. In the STS congenital database, among patients 12 to 30 years of age undergoing aortic valve surgery, 21% had placement of a mechanical valve, 18% had placement of a bioprosthetic valve, 30% had aortic valve repair, and 24% had a Ross procedure. In the future, the balance among these options may be altered by design improvements in prosthetic valves, alternatives to warfarin, the development of new patch materials for valve repair, and techniques to avoid Ross autograft failure.

  6. Frequency probabilistic analysis of a small break LOCA due to a power operated relief valve (PORV) for Angra-1 pre-TMI and post-TMI

    International Nuclear Information System (INIS)

    Onusic Junior, J.

    1986-01-01

    After the TMI event efforts were aimed towards improvements in the operational and administrative procedures related to the power operated relief valves (PORVs) in order to decrease the probability of a small-break loss-of-coolant accident (LOCA) caused by stuck-open power operated relief valve. This paper presents a frequency probabilistic analysis of a small break LOCA due to a stuck open PORV and safety valve to the Angra I nuclear power plant in operating conditions pre-TMI and post-TMI. (Author) [pt

  7. Aortic valve function after bicuspidization of the unicuspid aortic valve.

    Science.gov (United States)

    Aicher, Diana; Bewarder, Moritz; Kindermann, Michael; Abdul-Khalique, Hashim; Schäfers, Hans-Joachim

    2013-05-01

    Unicuspid aortic valve (UAV) anatomy leads to dysfunction of the valve in young individuals. We introduced a reconstructive technique of bicuspidizing the UAV. Initially we copied the typical asymmetry of a normal bicuspid aortic valve (BAV) (I), later we created a symmetric BAV (II). This study compared the hemodynamic function of the two designs of a bicuspidized UAV. Aortic valve function was studied at rest and during exercise in 28 patients after repair of UAV (group I, n = 8; group II, n = 20). There were no differences among the groups I and II with respect to gender, age, body size, or weight. All patients were in New York Heart Association class I. Six healthy adults served as control individuals. All patients were studied with transthoracic echocardiography between 4 and 65 months postoperatively. Systolic gradients were assessed by continuous wave Doppler while patients were at rest and exercising on a bicycle ergometer. Aortic regurgitation was grade I or less in all patients. Resting gradients were significantly elevated in group I compared with group II and control individuals (group I, peak 33.8 ± 7.8 mm Hg; mean 19.1 ± 5.4 mm Hg; group II, peak 15.8 ± 5.4, mean 8.2 ± 2.8 mm Hg; control individuals, peak 6.0 ± 1.6, mean 3.2 ± 0.8 mm Hg; p competence. A symmetric repair design leads to improved systolic aortic valve function at rest and during exercise. Copyright © 2013 The Society of Thoracic Surgeons. Published by Elsevier Inc. All rights reserved.

  8. Cardiac Injury After All-Terrain Vehicle Accidents in 2 Children and a Review of the Literature.

    Science.gov (United States)

    Ngo, Kimberly D; Pian, Phillip; Hanfland, Robert; Nichols, Christopher S; Merritt, Glenn R; Campbell, David; Ing, Richard J

    2016-07-01

    All-terrain vehicle (ATV) accidents leading to severe morbidity and mortality are common. At our institution, 2 children presented within weeks of each other after ATV accidents. Both children required cardiac valve surgery. The surgical management of these 2 children is discussed, and the literature is reviewed. On initial patient presentation, the diagnosis of a ruptured cardiac valve or ventricular septal defect (VSD) associated with these types of accidents is often delayed. We propose that patients presenting with evidence of high-energy blunt thoracic trauma after an ATV accident should undergo an electrocardiogram, cardiac enzyme assessment, and cardiac echocardiogram as part of the initial work-up to rule out significant myocardial injury.

  9. ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

    International Nuclear Information System (INIS)

    Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

    2002-01-01

    The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

  10. Database on aircraft accidents

    International Nuclear Information System (INIS)

    Nishio, Masahide; Koriyama, Tamio

    2012-09-01

    The Reactor Safety Subcommittee in the Nuclear Safety and Preservation Committee published the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' as the standard method for evaluating probability of aircraft crash into nuclear reactor facilities in July 2002. In response to the report, Japan Nuclear Energy Safety Organization has been collecting open information on aircraft accidents of commercial airplanes, self-defense force (SDF) airplanes and US force airplanes every year since 2003, sorting out them and developing the database of aircraft accidents for latest 20 years to evaluate probability of aircraft crash into nuclear reactor facilities. This year, the database was revised by adding aircraft accidents in 2010 to the existing database and deleting aircraft accidents in 1991 from it, resulting in development of the revised 2011 database for latest 20 years from 1991 to 2010. Furthermore, the flight information on commercial aircrafts was also collected to develop the flight database for latest 20 years from 1991 to 2010 to evaluate probability of aircraft crash into reactor facilities. The method for developing the database of aircraft accidents to evaluate probability of aircraft crash into reactor facilities is based on the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' described above. The 2011 revised database for latest 20 years from 1991 to 2010 shows the followings. The trend of the 2011 database changes little as compared to the last year's one. (1) The data of commercial aircraft accidents is based on 'Aircraft accident investigation reports of Japan transport safety board' of Ministry of Land, Infrastructure, Transport and Tourism. 4 large fixed-wing aircraft accidents, 58 small fixed-wing aircraft accidents, 5 large bladed aircraft accidents and 114 small bladed aircraft accidents occurred. The relevant accidents for evaluating

  11. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery managment concevtrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that 'active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk', and goes further in considering and formulating the key issue: 'The most fruitful path to follow in reducing risk even further is through the planning of accident management.' (author)

  12. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  13. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

    The definition and the multidimensionality aspects of accident management have been reviewed. The suggested elements in the development of a programme for severe accident management have been identified and discussed. The strategies concentrate on the two tiered approaches. Operative management utilizes the plant's equipment and operators capabilities. The recovery management concentrates on preserving the containment, or delaying its failure, inhibiting the release, and on strategies once there has been a release. The inspiration for this paper was an excellent overview report on perspectives on managing severe accidents in commercial nuclear power plants and extending plant operating procedures into the severe accident regime; and by the most recent publication of the International Nuclear Safety Advisory Group (INSAG) considering the question of risk reduction and source term reduction through accident prevention, management and mitigation. The latter document concludes that active development of accident management measures by plant personnel can lead to very large reductions in source terms and risk, and goes further in considering and formulating the key issue: The most fruitful path to follow in reducing risk even further is through the planning of accident management

  14. Persistence of airline accidents.

    Science.gov (United States)

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

  15. Conceptual design of a compact absolute valve for the ITER neutral beam injectors

    Energy Technology Data Exchange (ETDEWEB)

    Jones, Chris [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)], E-mail: chris.m.jones@jet.uk; Waldon, Chris; Martin, David; Watson, Mike [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Sonderegger, Kurt; Lenherr, Bruno [VAT Vakuumventile AG, CH-9469 Haag (Switzerland); Andrews, Ian; Mansbridge, Simon [VAT Vacuum Products Ltd., Edmund House, Rugby Road, Leamington Spa, Warwickshire CV32 6EL (United Kingdom)

    2009-06-15

    The reference design for the ITER neutral beam injectors incorporated a fast shutter to limit tritium migration to the injector vacuum enclosures. In 2005, a need for an 'absolute' isolation valve was identified to facilitate injector maintenance procedures and protect the system from an in-vessel ingress of coolant event (ICE). An outline concept for an all-metal seal valve was developed during 2006, in close cooperation with the Swiss valve manufacturer VAT. During the following year, it became apparent that the length of beamline available for the valve was significantly less than originally envisaged, resulting in a radical revision of the design concept. A casing length of 760 mm has been achieved by means of major changes to the casing structure, plate dimensions, pendulum mechanism and seal actuators. A concept for a seal protection system has been developed to prevent beam line contamination reaching the valve components and to protect the valve plate from surface heating by plasma radiation. The new design concept has been extensively validated by analysis, including a whole-system FE model of the valve.

  16. Real-time 3D visualization of cellular rearrangements during cardiac valve formation.

    Science.gov (United States)

    Pestel, Jenny; Ramadass, Radhan; Gauvrit, Sebastien; Helker, Christian; Herzog, Wiebke; Stainier, Didier Y R

    2016-06-15

    During cardiac valve development, the single-layered endocardial sheet at the atrioventricular canal (AVC) is remodeled into multilayered immature valve leaflets. Most of our knowledge about this process comes from examining fixed samples that do not allow a real-time appreciation of the intricacies of valve formation. Here, we exploit non-invasive in vivo imaging techniques to identify the dynamic cell behaviors that lead to the formation of the immature valve leaflets. We find that in zebrafish, the valve leaflets consist of two sets of endocardial cells at the luminal and abluminal side, which we refer to as luminal cells (LCs) and abluminal cells (ALCs), respectively. By analyzing cellular rearrangements during valve formation, we observed that the LCs and ALCs originate from the atrium and ventricle, respectively. Furthermore, we utilized Wnt/β-catenin and Notch signaling reporter lines to distinguish between the LCs and ALCs, and also found that cardiac contractility and/or blood flow is necessary for the endocardial expression of these signaling reporters. Thus, our 3D analyses of cardiac valve formation in zebrafish provide fundamental insights into the cellular rearrangements underlying this process. © 2016. Published by The Company of Biologists Ltd.

  17. Transient analysis for a system with a tilted disc check valve

    International Nuclear Information System (INIS)

    Jeung, Jaesik; Lee, Kyukwang; Cho, Daegwan

    2014-01-01

    Check valves are used to prevent reverse flow conditions in a variety of systems in nuclear power plants. When a check valve is closed by a reverse flow, the transient load can jeopardize the structural integrity on the piping system and its supports. It may also damage intended function of the in-line components even though the severity of the load differs and depends strongly on types of the check valves. To incorporate the transient load in the piping system, it is very important to properly predict the system response to transients such as a check valve closure accompanied by pump trip and to evaluate the system transient. The one-dimensional transient simulation codes such as the RELAP5/MOD3.3 and TRACE were used. There has not been a single model that integrates the two codes to handle the behavior of a tilted disc check valve, which is designed to mitigate check valve slams by shorting the travel of the disc. In this paper a model is presented to predict the dynamic motion of a tilted disc check valve in the transient simulation using the RELAP5/MOD3.3 code and the model is incorporated in a system transient analysis using control variables of the code. In addition, transient analysis for Essential Service Water (ESW) system is performed using the proposed model and the associated load is evaluated for the system. (author)

  18. Optothermally actuated capillary burst valve

    DEFF Research Database (Denmark)

    Eriksen, Johan; Bilenberg, Brian; Kristensen, Anders

    2017-01-01

    be burst by raising the temperature due to the temperature dependence of the fluid surface tension. We address individual valves by using a local heating platform based on a thin film of near infrared absorber dye embedded in the lid used to seal the microfluidic device [L. H. Thamdrup et al., Nano Lett...

  19. Spring valve for well completion

    Energy Technology Data Exchange (ETDEWEB)

    Gorbatov, P T

    1966-07-22

    A spring-loaded valve for well completion consists of a housing with a spring-loaded closing element. In order to protect the closing element from corrosion which might lower the pressure drop, the closing element is made in the form of a piston. It is tightly connected with sealing elements. The housing has orifices, overlapping the piston in the initial position.

  20. Hydraulic servo control spool valve

    Science.gov (United States)

    Miller, Donald M.

    1983-01-01

    A servo operated spool valve having a fixed sleeve and axially movable spool. The sleeve is machined in two halves to form a long, narrow tapered orifice slot across which a transverse wall of the spool is positioned. The axial position of the spool wall along the slot regulates the open orifice area with extreme precision.

  1. Guidelines for valves in tritium service

    International Nuclear Information System (INIS)

    Weaver, W.W.

    1994-01-01

    Some undesirable practices and misapplications that caused valve-related failures are examined, and future courses of action are recommended to avoid repetition of these events. Desirable valve characteristics and practices that should be considered when selecting valves for use in tritium service are also discussed. Supporting logic for the desirability of these features is presented by discussing the mechanisms of valve degradation followed by examples of related events. Desirable valve and system features and operational actions are grouped into two categories: strongly recommended and recommended. 13 refs., 1 fig

  2. Valve assembly having remotely replaceable bearings

    International Nuclear Information System (INIS)

    Johnson, E.R.; Tanner, D.E.

    1980-01-01

    A valve assembly having remotely replaceable bearings is disclosed wherein a valve disc is supported within a flow duct for rotation about a pair of axially aligned bearings, one of which is carried by a spindle received within a diametral bore in the valve disc, and the other of which is carried by a bearing support block releasably mounted on the duct circumferentially of an annular collar on the valve disc coaxial with its diametrical bore. The spindle and bearing support block are adapted for remote removal to facilitate servicing or replacement of the valve disc support bearings

  3. Social impact of accidents

    International Nuclear Information System (INIS)

    Kuroda, Isao

    1997-01-01

    There is the quite big difference between technological risk and social risk feeling. Various biases of social and sensational factors on accidents must be considered to recognize this difference. 'How safe is safe enough' is the perpetual thema concerning with not only technology but also sociology. The safety goal in aircraft design and how making effort to improve the present safety status in civil jet aircrafts is discussed as an example of social risk allowance. INSAG under IAEA started to discuss the safety culture after Chernobyl nuclear power plant accident on 1986. Safety culture and risk communication are the most important procedures to relieve the social impact for accidents. (author)

  4. Severe accident behavior

    International Nuclear Information System (INIS)

    Denning, R.S.

    1986-01-01

    The purpose of this paper is to provide an overview of severe accident behavior. The term source term is defined and a brief history of the regulatory use of source term is presented. The processes in severe accidents in light water reactors are described with particular emphasis on the relationships between accident thermal-hydraulics and chemistry. Those factors which have the greatest impact on predicted source terms are identified. Design differences between plants that affect source term estimation are also described. The principal unresolved issues are identified that are the focus of ongoing research and debate in the technical community

  5. Management of accident risks

    International Nuclear Information System (INIS)

    Compes, P.C.

    1987-01-01

    The example of the Chernobyl accident and the statistics of the occurrence of accidents make clear the threat to humanity, if one cannot guarantee successful accident prevention in the use and distribution of the projects aimed at. The science of safety, as it is known in the Wuppertal model, makes its contribution to this vital task for the human community. It makes it necessary to create the essential dates and concepts, the methods, principles and techniques based on them and the associated instrumentation. (DG) [de

  6. Chernobyl accident and Denmark

    International Nuclear Information System (INIS)

    1986-12-01

    The report describes the Chernobyl accident and its consequences for Denmark in particular. It was commissioned by The Secretary of State for the Environment. The event at the accident site, the release and dispersal of radioactive substances into the atmosphere and over Europe, is described. A discussion of the Danish organisation for nuclear emergencies, how it was activated and adapted to the actual situation, is given. A comprehensive description of the radiological contamination in Denmark following the accident and the estimated health effects, is presented. The situation in other European countries is mentioned. (author)

  7. Multifunctional four-port directional control valve constructed from logic valves

    International Nuclear Information System (INIS)

    Lisowski, E.; Czyżycki, W.; Rajda, J.

    2014-01-01

    Highlights: • Directional valve with standard ISO 440-08 has been constructed from logic valves. • Only one innovative valve may replace whole family of the standard valves. • CFD analysis and bench tests of the innovative valve has been carried. • Parameters of the innovative valve are equaling or surpassing the standard ones. • The innovative valve has additional possibilities of pressure and flow control. - Abstract: The paper refers to four-port solenoid pilot operated valves, which are subplate mounted in a hydraulic system in accordance with the ISO 4401 standard. Their widespread use in many machines and devices causes a continuing interest in the development of their design by both the scientific centers and the industry. This paper presents an innovative directional control valve based on the use of logic valves and a methodology followed for the design of it by using Solid Edge CAD and ANSYS/Fluent CFD software. The valve design methodology takes into account the need to seek solutions that minimize flow resistance through the valve. For this purpose, the flow paths are prepared by means of CAD software and pressure-flow curves are determined as a result of CFD analysis. The obtained curves are compared with the curves available in the catalogs of spool type directional control valves. The new solution allows to replace the whole family of spool type four-port directional control valves by one valve built of logic valves. In addition, the innovative directional control valve provides leak-proof shutting the flow paths off and also it can control flow rate and even pressure of working liquid. A prototype of the valve designed by the presented method has been made and tested on the test bench. The results quoted in the paper confirm that the developed logic type directional control valve is able to meet all designed connection configurations, and the obtained pressure-flow curves show very good conformity with the results of CFD analysis

  8. [Tricuspid valve insufficiency: what should be done?].

    Science.gov (United States)

    von Segesser, L K; Stauffer, J C; Delabays, A; Chassot, P G

    1998-12-01

    Tricuspid regurgitation is relatively common. Due to the progress made in echocardiography, its diagnosis is in general made readily and in reliable fashion. Basically one has to distinguish between functional tricuspid valve regurgitation due to volume and/or pressure overload of the right ventricle with intact valve structures versus tricuspid valve regurgitation due to pathologic valve structures. The clear identification of the regurgitation mechanism is of prime importance for the treatment. Functional tricuspid valve regurgitation can often be improved by medical treatment of heart failure, and eventually a tricuspid valve plasty can solve the problem. However, the presence of pathologic tricuspid valve structures makes in general more specific plastic surgical procedures and even prosthetic valve replacements necessary. A typical example for a structural tricuspid valve regurgitation is the case of a traumatic papillary muscle rupture. Due to the sudden onset, this pathology is not well tolerated and requires in general surgical reinsertion of the papillary muscle. In contrast, tricuspid valve regurgitation resulting from chronic pulmonary embolism with pulmonary artery hypertension, can be improved by pulmonary artery thrombendarteriectomy and even completely cured with an additional tricuspid annuloplasty. However, tricuspid regurgitations due to terminal heart failure are not be addressed with surgery directed to tricuspid valve repair or replacement. Heart transplantation, dynamic cardiomyoplasty or mechanical circulatory support should be evaluated instead.

  9. Mechanical versus bioprosthetic aortic valve replacement.

    Science.gov (United States)

    Head, Stuart J; Çelik, Mevlüt; Kappetein, A Pieter

    2017-07-21

    Mechanical valves used for aortic valve replacement (AVR) continue to be associated with bleeding risks because of anticoagulation therapy, while bioprosthetic valves are at risk of structural valve deterioration requiring reoperation. This risk/benefit ratio of mechanical and bioprosthetic valves has led American and European guidelines on valvular heart disease to be consistent in recommending the use of mechanical prostheses in patients younger than 60 years of age. Despite these recommendations, the use of bioprosthetic valves has significantly increased over the last decades in all age groups. A systematic review of manuscripts applying propensity-matching or multivariable analysis to compare the usage of mechanical vs. bioprosthetic valves found either similar outcomes between the two types of valves or favourable outcomes with mechanical prostheses, particularly in younger patients. The risk/benefit ratio and choice of valves will be impacted by developments in valve designs, anticoagulation therapy, reducing the required international normalized ratio, and transcatheter and minimally invasive procedures. However, there is currently no evidence to support lowering the age threshold for implanting a bioprosthesis. Physicians in the Heart Team and patients should be cautious in pursuing more bioprosthetic valve use until its benefit is clearly proven in middle-aged patients. Published on behalf of the European Society of Cardiology. All rights reserved. © The Author 2017. For permissions, please email: journals.permissions@oup.com.

  10. Acoustic valve leak detection in nuclear plants

    International Nuclear Information System (INIS)

    Dimmick, J.G.; Dickey, J.W.

    1983-01-01

    Internal valve leakage is a hidden energy loss and can cause or prolong a forced outage. Recent advances in acoustic detection of internal valve leakage have reduced piping system maintenance costs, unnecessary downtime, and energy waste. Extremely short payback periods have been reported by plants applying this technology to preventive maintenance, troubleshooting, energy conservation and outage planning. Sensors temporarily attached to the outside of valves and connected to the instruments detect ultrasonic acoustic emissions which are characteristic of internal valve leakage. Since the sensors are attached to the outside of the valves, the time and expense of dismantling the valves or removing them from the systems are eliminated. This paper describes the instrumentation and specific applications to nuclear plant valves, including independent verification of initial findings. Guidelines for potential users, including instrumentation selection, training requirements, application planning, and the choice of in-house versus contract services are discussed

  11. Operating experience feedback report -- Pressure locking and thermal binding of gate valves

    International Nuclear Information System (INIS)

    Hsu, C.

    1993-03-01

    The potential for valve inoperability caused by pressure locking and thermal binding has been known for many years in the nuclear industry. Pressure locking or thermal binding is a common-mode failure mechanism that can prevent a gate valve from opening, and could render redundant trains of safety systems or multiple safety systems inoperable. In spite of numerous generic communications issued in the past by the Nuclear Regulatory Commission (NRC) and industry, pressure locking and thermal binding continues to occur to gate valves installed in safety-related systems of both boding water reactors (BWRs) and pressurized water reactors (PWRs). The generic communications to date have not led to effective industry action to fully identify, evaluate, and correct the problem. This report provides a review of operating events involving these failure mechanisms. As a result of this review this report: (1) identifies conditions when the failure mechanisms have occurred, (2) identifies the spectrum of safety systems that have been subjected to the failure mechanisms, and (3) identifies conditions that may introduce the failure mechanisms under both normal and accident conditions. On the basis of the evaluation of the operating events, the Office for Analysis and Evaluation of Operational Data (AEOD) of the NRC concludes that the binding problems with gate valves are an important safety issue that needs priority NRC and industry attention. This report also provides AEOD's recommendation for actions to effectively prevent the occurrence of valve binding failures

  12. Replacement screws valve operating under Trunnion; Substituicao de parafusos de valvulas Trunnion em regime de operacao

    Energy Technology Data Exchange (ETDEWEB)

    Souza Netto, Charles de; Santos, Rogerio Andre Zolin dos; Arnhold, Diego [Companhia de Gas do Estado do Rio Grande do Sul (SULGAS), Porto Alegre, RS (Brazil); Jacques, Rodrigo das Neves [Guidotti e Vieira Manutencao Industrial Ltda., Canoas, RS (Brazil)

    2012-07-01

    The report shows the process created for the substitution and extraction of bearing screws of the Trunnion valves, in operation. The methodology was developed at the 'Companhia de Gas do Estado do Rio Grande do Sul - SULGAS', with the objective of avoiding failure emergency situations, and or sudden breaking of the screws of fixation of the lid of the inferior bearing of the Trunnion valves. it is a preventive process of substitution of these screws, that after a great period of use in atmospheres with high potential of oxidation present structural failure. The breaking of these components creates a leaking process by the inferior lid of the valves, fact that is intended to be avoided with the application of the technical procedure of this report, guaranteeing the integrity of the valves that are vital components for the continuous operation of the gas pipe line. (author)

  13. Experimental And Numerical Investigation Of The Flow Analysis Of The Water-Saving Safety Valve

    Directory of Open Access Journals (Sweden)

    Muhammed Safa Kamer

    2015-08-01

    Full Text Available Abstract In this study an auto-mechanical safety valve was designed and manufactured in order to prevent possible wastage of water and water raid after instantaneous water cuts during water usage in places where water use is widespread. Safety valve is activated and it switches off the line when water is cut off when mains pressure is equal to atmospheric pressure and as it does not allow water to pass when it comes back it saves water and prevents the formation of raids. An experiment set was conducted in order to measure the pressure drop between the inlet and outlet of the safety valve and it was found that with the increased flow rate the pressure drop increases. The three-dimensional flow analysis of the safety valve was carried out with Ansys-Fluent software package and the results obtained were compared with experimental data and a good harmony was achieved.

  14. Advances in small zero-leak valves point to better nuclear power-plant reliability

    Energy Technology Data Exchange (ETDEWEB)

    Eacott, K B; Kin, J C; Hotta, Y [Dresser Japan, Ltd.

    1978-04-01

    In the selection of small valves less than two inches used for nuclear power plants, sufficient consideration must be given to the reliability to radioactive material, the easy operability, and the significant function, especially zero leak. These valves are classified into bellows and diaphragm seal types which must satisfy zero leak, 4000 cycles life test and good maintainability. Welded bellows, formed bellows, and metal diaphragms are actually used for these requirements. The construction of these types are shown. The requirements and principal specifications for these small valves are explained, and some examples are given. These zero leak valves are installed in reactor coolant loop system, borated water from B. A. system, pressurizer instrument system, containment spray system, high head system and off gas system for PWRS, and main steam line system, diesel generator cooling water system, re-circulation system, clean up water system, etc. for BWRS.

  15. Condition monitoring of a check valve for nuclear power plants by means of acoustic emission technique

    International Nuclear Information System (INIS)

    Lee, M. R.; Lee, J. H.; Kim, J. T.; Kim, J. S.; Luk, V. K.

    2003-01-01

    This work performed in support of the International Nuclear Energy Research Institute (INERI) program, which was to develop and demonstrate advanced sensor and computational technology for on-line monitoring of the condition of components, structures, and systems in advanced and next-generation nuclear power plants (NPPs). This primary object of this work is to investigate advanced condition monitoring systems based on acoustic emission detection that can provide timely detection of check valve degeneration and service aging so that maintenance/replacement could be preformed prior to loss safety function. The research is focused on the capability of AE technique to provide diagnostic information useful in determining check valve aging and degradation check valve failure and undesirable operating modes. This work also includes the investigation and adaptation of several advanced sensor technologies such as accelerometer and advanced ultrasonic technique. In addition, this work will develop advanced sophisticated signal processing, noise reduction, and pattern recognition techniques and algorithms from check valve degradation.

  16. Accident resistant transport container

    Science.gov (United States)

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  17. Accident resistant transport container

    International Nuclear Information System (INIS)

    Andersen, J.A.; Cole, J.K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident

  18. Big nuclear accidents

    International Nuclear Information System (INIS)

    Marshall, W.; Billingon, D.E.; Cameron, R.F.; Curl, S.J.

    1983-09-01

    Much of the debate on the safety of nuclear power focuses on the large number of fatalities that could, in theory, be caused by extremely unlikely but just imaginable reactor accidents. This, along with the nuclear industry's inappropriate use of vocabulary during public debate, has given the general public a distorted impression of the risks of nuclear power. The paper reviews the way in which the probability and consequences of big nuclear accidents have been presented in the past and makes recommendations for the future, including the presentation of the long-term consequences of such accidents in terms of 'loss of life expectancy', 'increased chance of fatal cancer' and 'equivalent pattern of compulsory cigarette smoking'. The paper presents mathematical arguments, which show the derivation and validity of the proposed methods of presenting the consequences of imaginable big nuclear accidents. (author)

  19. Accidents in perspective

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1989-01-01

    The nuclear industry perspective and the public perspective on big nuclear accidents and leukaemia near nuclear sites are discussed. The industry perspective is that big accidents are so unlikely as to be virtually impossible and that leukaemia is not specifically associated with nuclear installations. Clusters of cancer with statistical significance occur in major cities. The public perspective is coloured by a prejudice and myth: the fear of radiation. The big nuclear accident is seen therefore as much more unacceptable than any other big accident. Risks associated with Sizewell-B nuclear station and the liquid gas depot at Canvey Island are discussed. The facts and figures are presented as tables and graphs. Given conflicting interpretations of the leukaemia problem the public inclines towards the more pessimistic view. (author)

  20. Boating Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  1. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  2. Use of intraventricular ribbon gauze to reduce particulate emboli during aortic valve replacement

    Directory of Open Access Journals (Sweden)

    Loubani Mahmoud

    2006-11-01

    Full Text Available Abstract Background The incidence of cerebrovascular accidents following aortic valve surgery remains a devastating complication. The aim of this study was to determine the number of potential embolic material arising during aortic valve replacement and to examine the efficacy of using ribbon gauze in the left ventricle during removal of the native valve and decalcification of the aortic annulus. Methods Ribbon gauze was inserted into the left ventricular cavity prior to aortic valve excision in an unselected, prospectively studied series of 30 patients undergoing aortic valve replacement. A further 30 lengths of ribbon gauze were soaked in the pericardiotomy blood of the same patients and all were subjected to histological analysis. Results The median number of tissue fragments from the aortic valve replacement group was significantly higher than in the control group 5 (0–18 versus 0 (0–1 (p = 3.6 × 10-5. The size of tissue fragments varied between 0.1 and 9.0 mm with a mean of 0.61 ± 1.12 mm and a median of 0.2 mm. There was a significantly higher number of tissue fragments associated with patients having surgery for aortic stenosis when compared with patients who had aortic regurgitation with median of 5 (0–18 versus 0 (0–3 (p = 0.8 × 10-3. Conclusion Significant capture of particulate debris by the intraventricular ribbon gauze suggests that the technique of left ventricular ribbon gauze insertion during aortic valve excision has merit.

  3. [Aortic valve insufficiency due to rupture of the cusp in a patient with multiple trauma].

    Science.gov (United States)

    Vidmar, J; Brilej, D; Voga, G; Kovacic, N; Smrkolj, V

    2003-06-01

    Lesions of the heart valve caused by blunt chest trauma is rare, but when it does occur it can significantly injure the patient. On the basis of autopsy studies, research shows that heart valves are injured in less than 5% of patients who have died due to impact thoracic trauma. Among the heart valves, the aortic valve is the most often lacerated, which has been proved by relevant autopsy and clinical studies. Aortic valve lesions can be the only injury, but it is possible that additional heart or large vessel injuries are also present (myocardial contusion, rupture of the atrial septum, aortic rupture, rupture of the left common carotid artery). The force that causes such an injury is often great and often causes injuries to other organs and organ systems. In a multiple trauma patient, it is very important to specifically look for heart-related injuries because it is possible that they may be overlooked or missed by the surgeon, because of other obvious injuries. We describe the case of a 41-year-old man with multiple trauma who was diagnosed with aortic valve insufficiency due to rupture of the left coronary cusp 6 weeks after a road accident. Valvuloplasty was performed. Seven years later the patient is free of symptoms and is in good physical condition. Echocardiography showed normal dimensions of the heart chambers, a normal thickness of the heart walls, and normal systolic and diastolic function of the left ventricle. Heart valves are morphologically normal, and only an unimportant aortic insufficiency was noticed by echocardiography.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Analysis of the RBMK-1500 type reactor emergency core cooling system behavior, taking into account the specified hydraulic characteristics of fast acting motor valves

    International Nuclear Information System (INIS)

    Kaliatka, A.; Ognerubov, V.; Adomavicius, A.; Ziedelis, S.

    2005-01-01

    During the accident analysis of nuclear power plants, reliability and uncertainty of results depends on adequateness of mathematical models of main elements and phenomena in systems important to safety. The best way for qualification of these models is collation with relevant experimental data. However, at the case of lack of such data modern computational fluid dynamics codes can be used for this purpose. This paper presents the results of an attempt to specify the hydraulic characteristics of the fast acting motor valves as well as to demonstrate the impact of these characteristics to transient processes in emergency core cooling system of the RBMK-1500 type reactor. For these purposes the finite element model of fast acting motor valve was developed and analyzed, using two separate computational fluid dynamics codes in parallel: CFX5 and COSMOS/FLOWORKS. Both all main design particularities and changes of flow structure during valve opening (closure) process were taken into account. It was demonstrated, that the obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate substantially differ from those commonly used in thermal-hydraulic calculations of nuclear reactors. This difference is extremely big at the square one of the valve opening process, when the value of the valve hydraulic resistance is most important to flow of coolant channelized to the group distribution header. The series of thermal-hydraulic calculations of the maximum design-basis accident initiated by full break of main circulation pump pressure header were performed. The obtained dependencies of changes of hydraulic loss coefficient in respect of relative valve opening (closure) rate as well as those commonly used in thermal-hydraulic code RELAP5 were used. The results of calculations show, that in the initial stage of accident flow of coolant going from emergency core cooling system via fast acting motor valves to group distribution

  6. Transcatheter aortic valve prosthesis surgically replaced 4 months after implantation

    DEFF Research Database (Denmark)

    Thyregod, Hans Gustav; Lund, Jens Teglgaard; Engstrøm, Thomas

    2010-01-01

    Transcatheter aortic valve implantation is a new and rapidly evolving treatment option for high-risk surgical patients with degenerative aortic valve stenosis. Long-term results with these new valve prostheses are lacking, and potential valve dysfunction and failure would require valve replacemen....... We report the first case of surgical valve replacement in a patient with a dysfunctional transcatheter-implanted aortic valve prosthesis 4 months after implantation....

  7. Indonesian Sea Accident Analysis (Case Study From 2003 – 2013)

    Science.gov (United States)

    Arya Dewanto, Y.; Faturachman, D.

    2018-03-01

    There are so many accidents in sea transportation in Indonesia. Most of the accidents happen because of low concern aspects of the safety and security of the crew. In sailing, a man as transport users to interact with the ship and the surrounding environment (including other ships, cruise lines, ports, and the situation of local conditions). These interactions are sometimes very complex and related to various aspects of. Aware of the multiplicity of aspects related to the third of these factors, seeking the safety of cruise through a reduction in the number of accidents and the risk of death and serious injuries due to accidents and goods transported is certainly not enough attempted through mono-sector approach, but rather takes a multi-sector approach to the efforts. In this paper, we described the Indonesian Sea Transportation accident analysis for eleven years divided into four items: total of ship accident type, ship accident factor, total of casualties, region of ship accidents. All data founded from Marine Court (Mahkamah Pelayaran). From that 4 items we can find Indonesia Sea Accident Analysis from 2003-2013.

  8. The 1986 Chernobyl accident; Der Unfall von Tschernobyl 1986

    Energy Technology Data Exchange (ETDEWEB)

    Kerner, Alexander; Stueck, Reinhard; Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching bei Muenchen, Koeln (Germany). Bereich Reaktorsicherheitsanalysen; Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Koeln (Germany)

    2011-02-15

    April 26, 2011 marks the 25th anniversary of the Chernobyl reactor accident, the worst incident in the history of the peaceful utilization of nuclear power. While investigations of the course of events and the causes of the accident largely present a uniform picture, descriptions still vary widely when it comes to the impact on the population and the environment. This treatment of the Chernobyl accident constitutes a summary of facts about the initiation of the accident and the sequence of events that followed. In addition, measures are described which were taken to exclude any repetition of a disaster of this kind. The health consequences and the socio-economic impact of the accident are not discussed in any detail. The first section contains an introduction and an overview of the Soviet RBMK (Chernobyl) reactor line. In section 2, fundamental characteristics of this special type of reactor, which was exclusively built in the former Soviet Union, are discussed. This information is necessary to understand the sequence of accident events and provides an answer to the frequent question whether that accident could be transferred to reactors in this country. The third section outlines the history of the accident caused ultimately by a commissioning test never performed before. The section is completed by a brief description of radiological releases and the state of the plant after the accident when entombed in the ''sarcophagus.'' The different causes are then summarized and the modifications afterwards made to RBMK reactors are outlined. (orig.)

  9. Alternative evacuation strategies for nuclear power accidents

    International Nuclear Information System (INIS)

    Hammond, Gregory D.; Bier, Vicki M.

    2015-01-01

    In the U.S., current protective-action strategies to safeguard the public following a nuclear power accident have remained largely unchanged since their implementation in the early 1980s. In the past thirty years, new technologies have been introduced, allowing faster computations, better modeling of predicted radiological consequences, and improved accident mapping using geographic information systems (GIS). Utilizing these new technologies, we evaluate the efficacy of alternative strategies, called adaptive protective action zones (APAZs), that use site-specific and event-specific data to dynamically determine evacuation boundaries with simple heuristics in order to better inform protective action decisions (rather than relying on pre-event regulatory bright lines). Several candidate APAZs were developed and then compared to the Nuclear Regulatory Commission’s keyhole evacuation strategy (and full evacuation of the emergency planning zone). Two of the APAZs were better on average than existing NRC strategies at reducing either the radiological exposure, the population evacuated, or both. These APAZs are especially effective for larger radioactive plumes and at high population sites; one of them is better at reducing radiation exposure, while the other is better at reducing the size of the population evacuated. - Highlights: • Developed framework to compare nuclear power accident evacuation strategies. • Evacuation strategies were compared on basis of radiological and evacuation risk. • Current strategies are adequate for smaller scale nuclear power accidents. • New strategies reduced radiation exposure and evacuation size for larger accidents

  10. Occupational Accidents And Preventive Measures

    CERN Document Server

    Fassnacht, V

    2006-01-01

    This report presents the 2005 statistics concerning occupational accidents involving members of the CERN personnel and contractors' personnel. It sets out the accident frequency and severity rates and provides a breakdown of accidents by cause and injury. It also contains a summary analysis of the most serious accidents and the associated recommendations.

  11. Accidents with sulfuric acid

    OpenAIRE

    Rajković Miloš B.

    2006-01-01

    Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eigh...

  12. The Chernobyl accident

    International Nuclear Information System (INIS)

    Berg, J.O.; Christensen, G.; Lingjaerde, R.; Smidt Olsen, H.; Wethe, P.I.

    1986-10-01

    In connection with the Chernobyl accident the report gives a description of the technical features of importance to the accident, the course of events, and the estimated health hazards in the local environment. Dissimilarities in western and Sovjet reactor safety philosophy are dealt with, as well as conceivable concequences in relation to technology and research in western nuclear power programmes. Results of activity level measurements of air and foodstuff, made in Norway by Institute for Energy Technology, are given

  13. Development of Long-Lifetime Pulsed Gas Valves for Pulsed Electric Thrusters

    Science.gov (United States)

    Burkhardt, Wendel M.; Crapuchettes, John M.; Addona, Brad M.; Polzin, Kurt A.

    2015-01-01

    It is advantageous for gas-fed pulsed electric thrusters to employ pulsed valves so propellant is only flowing to the device during operation. The propellant utilization of the thruster will be maximized when all the gas injected into the thruster is acted upon by the fields produced by the electrical pulse. Gas that is injected too early will diffuse away from the thruster before the electrical pulse can act to accelerate the propellant. Gas that is injected too late will miss being accelerated by the already-completed electrical pulse. As a consequence, the valve must open quickly and close equally quickly, only remaining open for a short duration. In addition, the valve must have only a small amount of volume between the sealing body and the thruster so the front and back ends of the pulse are as coincident as possible with the valve cycling, with very little latent propellant remaining in the feed lines after the valve is closed. For a real mission of interest, a pulsed thruster can be expected to pulse at least 10(exp 10) - 10(exp 11) times, setting the range for the number of times a valve must open and close. The valves described in this paper have been fabricated and tested for operation in an inductive pulsed plasma thruster (IPPT) for in-space propulsion. In general, an IPPT is an electrodeless space propulsion device where a capacitor is charged to an initial voltage and then discharged, producing a high-current pulse through a coil. The field produced by this pulse ionizes propellant, inductively driving current in a plasma located near the face of the coil. Once the plasma is formed, it can be accelerated and expelled at a high exhaust velocity by the electromagnetic Lorentz body force arising from the interaction of the induced plasma current and the magnetic field produced by the current in the coil. The valve characteristics needed for the IPPT application require a fast-acting valve capable of a minimum of 10(exp 10) valve actuation cycles. Since

  14. Method of effecting fast turbine valving for improvement of power system stability

    International Nuclear Information System (INIS)

    Park, R.H.

    1981-01-01

    As a improved way of effecting fast valving of turbines of power system steam-electric generating units for the purpose of improving the stability of power transmission over transmission circuits to which their generators make connection, when stability is threatened by line faults and certain other stability endangering events, the heretofore employed and/or advocated practice of automatically closing intercept valves at fastest available closing speed in response to a fast valving signal, and thereafter automatically fully reopening them in a matter of seconds, is modified by providing to reopen the valves only partially to and thereafter retain them at a preset partially open position. For best results the process of what can be termed sustained partial reopening is so effected as to result in its completion within a fraction of a second following the peak of the first forward swing of the generator rotor. Control valves may be either held open, or automatically fully or partly closed and thereafter fully opened in a preprogrammed manner, or automatically moved to and thereafter held in a partly closed position, by means of a preprogrammed process of repositioning in which the valves may optionally be first fully or partly closed and thereafter partly reopened. Avoidance of discharge of steam through high pressure safety valves can be had with use of suitably controlled power operated valves that discharge steam to the condenser or to atmosphere. Where there is an intermediate pressure turbine that is supplied with superheated steam, use of sustained partial control valve closure, if employed, is supplemented by provision for reduction of rate of heat release within the steam generator in order to protect the reheater from overheating. As a way to restrict increase of reheat pressure of fossil fuel installations, and to minimize increase in the msr (Moisture separator-reheater) pressure of nuclear units, provision is optionally made of normally closed by-pass v

  15. Accident and emergency management

    International Nuclear Information System (INIS)

    Andersen, V.; Moellenbach, K.; Heinonen, R.; Jakobsson, S.; Kukko, T.; Berg, Oe.; Larsen, J.S.; Westgaard, T.; Magnusson, B.; Andersson, H.; Holmstroem, C.; Brehmer, B.; Allard, R.

    1988-06-01

    There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

  16. Accident management information needs

    International Nuclear Information System (INIS)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs

  17. Accident management information needs

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1990-04-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs.

  18. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  19. Database on aircraft accidents

    International Nuclear Information System (INIS)

    Nishio, Masahide; Koriyama, Tamio

    2013-11-01

    The Reactor Safety Subcommittee in the Nuclear Safety and Preservation Committee published 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' as the standard method for evaluating probability of aircraft crash into nuclear reactor facilities in July 2002. In response to this issue, Japan Nuclear Energy Safety Organization has been collecting open information on aircraft accidents of commercial airplanes, self-defense force (SDF) airplanes and US force airplanes every year since 2003, sorting out them and developing the database of aircraft accidents for the latest 20 years to evaluate probability of aircraft crash into nuclear reactor facilities. In this report the database was revised by adding aircraft accidents in 2011 to the existing database and deleting aircraft accidents in 1991 from it, resulting in development of the revised 2012 database for the latest 20 years from 1992 to 2011. Furthermore, the flight information on commercial aircrafts was also collected to develop the flight database for the latest 20 years from 1992 to 2011 to evaluate probability of aircraft crash into reactor facilities. The method for developing the database of aircraft accidents to evaluate probability of aircraft crash into reactor facilities is based on the report 'The criteria on assessment of probability of aircraft crash into light water reactor facilities' described above. The 2012 revised database for the latest 20 years from 1992 to 2011 shows the followings. The trend of the 2012 database changes little as compared to the last year's report. (1) The data of commercial aircraft accidents is based on 'Aircraft accident investigation reports of Japan transport safety board' of Ministry of Land, Infrastructure, Transport and Tourism. The number of commercial aircraft accidents is 4 for large fixed-wing aircraft, 58 for small fixed-wing aircraft, 5 for large bladed aircraft and 99 for small bladed aircraft. The relevant accidents

  20. Application of ceramics to the sliding seat of valve bridge; Valve bridge yodobu eno ceramics tekiyo

    Energy Technology Data Exchange (ETDEWEB)

    Matsui, T; Ono, T [Mitsubishi Motors Corp., Tokyo (Japan)

    1997-10-01

    For use in the valve train, using an OHV (over head valve) configuration. of a 4 valve diesel engine for trucks and buses; we developed a valve bridge, a component of a valve train, with a ceramic head that is made of silicon nitride(Si3N4) in contact with a rocker arm in order to reduce cost and improve wear resistance for further diesel engine emissions regulations. In order to evaluate the effect of this valve bridge, RIG tests and durability tests on actual engines were carried out. 7 figs., 2 tabs.

  1. Historical aspects of radiation accidents

    International Nuclear Information System (INIS)

    Mettler, F.A. Jr.; Ricks, R.C.

    1990-01-01

    Radiation accidents are extremely rare events; however, the last two years have witnessed the largest radiation accidents in both the eastern and western hemispheres. It is the purpose of this chapter to review how radiation accidents are categorized, examine the temporal changes in frequency and severity, give illustrative examples of several types of radiation accidents, and finally, to describe the various registries for radiation accidents

  2. 7 CFR 58.219 - High pressure pumps and lines.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 3 2010-01-01 2010-01-01 false High pressure pumps and lines. 58.219 Section 58.219....219 High pressure pumps and lines. High pressure lines may be cleaned-in-place and shall be of such construction that dead ends, valves and the high pressure pumps can be disassembled for hand cleaning. The high...

  3. Performance of balanced bellows safety relief valves

    International Nuclear Information System (INIS)

    Lai, Y.S.

    1992-01-01

    By the nature of its design, the set point and lift of a conventional spring loaded safety relief valve are sensitive to back pressure. One way to reduce the adverse effects of the back pressure on the safety relief valve function is to install a balanced bellows in a safety relief valve. The metallic bellows has a rather wide range of manufacturing tolerance which makes the design of the bellows safety relief valve very complicated. The state-of-the-art balanced bellows safety relief valve can only substantially minimize, but cannot totally eliminate the back pressure effects on its set point and relieving capacity. Set point change is a linear function of the back pressure to the set pressure ratio. Depending on the valve design, the set point correction factor can be either greater or smaller than unity. There exists an allowable back pressure and critical back pressure for each safety relief valve. When total back pressure exceeds the R a , the relieving capacity will be reduced mainly resulting from the valve lift being reduced by the back pressure and the capacity reduction factor should be applied in valve sizing. Once the R c is exceeded, the safety relief valve becomes unstable and loses its over pressure protection capability. The capacity reduction factor is a function of system overpressure, but their relationship is non-linear in nature. (orig.)

  4. Traumatic Mitral Valve and Pericardial Injury

    Directory of Open Access Journals (Sweden)

    Nissar Shaikh

    2013-01-01

    Full Text Available Cardiac injury after blunt trauma is common but underreported. Common cardiac trauma after the blunt chest injury (BCI is cardiac contusion; it is very rare to have cardiac valve injury. The mitral valve injury during chest trauma occurs when extreme pressure is applied at early systole during the isovolumic contraction between the closure of the mitral valve and the opening of the aortic valve. Traumatic mitral valve injury can involve valve leaflet, chordae tendineae, or papillary muscles. For the diagnosis of mitral valve injury, a high index of suspicion is required, as in polytrauma patients, other obvious severe injuries will divert the attention of the treating physician. Clinical picture of patients with mitral valve injury may vary from none to cardiogenic shock. The echocardiogram is the main diagnostic modality of mitral valve injuries. Patient’s clinical condition will dictate the timing and type of surgery or medical therapy. We report a case of mitral valve and pericardial injury in a polytrauma patient, successfully treated in our intensive care unit.

  5. Simple Check Valves for Microfluidic Devices

    Science.gov (United States)

    Willis, Peter A.; Greer, Harold F.; Smith, J. Anthony

    2010-01-01

    A simple design concept for check valves has been adopted for microfluidic devices that consist mostly of (1) deformable fluorocarbon polymer membranes sandwiched between (2) borosilicate float glass wafers into which channels, valve seats, and holes have been etched. The first microfluidic devices in which these check valves are intended to be used are micro-capillary electrophoresis (microCE) devices undergoing development for use on Mars in detecting compounds indicative of life. In this application, it will be necessary to store some liquid samples in reservoirs in the devices for subsequent laboratory analysis, and check valves are needed to prevent cross-contamination of the samples. The simple check-valve design concept is also applicable to other microfluidic devices and to fluidic devices in general. These check valves are simplified microscopic versions of conventional rubber- flap check valves that are parts of numerous industrial and consumer products. These check valves are fabricated, not as separate components, but as integral parts of microfluidic devices. A check valve according to this concept consists of suitably shaped portions of a deformable membrane and the two glass wafers between which the membrane is sandwiched (see figure). The valve flap is formed by making an approximately semicircular cut in the membrane. The flap is centered over a hole in the lower glass wafer, through which hole the liquid in question is intended to flow upward into a wider hole, channel, or reservoir in the upper glass wafer. The radius of the cut exceeds the radius of the hole by an amount large enough to prevent settling of the flap into the hole. As in a conventional rubber-flap check valve, back pressure in the liquid pushes the flap against the valve seat (in this case, the valve seat is the adjacent surface of the lower glass wafer), thereby forming a seal that prevents backflow.

  6. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  7. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    International Nuclear Information System (INIS)

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

  8. Organic evaporator steam valve failure

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1992-01-01

    Defense Waste Processing Facility (DWPF) Technical has requested an analysis of the capacity of the Organic Evaporator (OE) condenser (OEC) be performed to determine its capability in the case where the OE steam flow control valve fails open. Calculations of the OE boilup and the OEC heat transfer coefficient indicate the OEC will have more than enough capacity to remove the heat at maximum OE boilup. In fact, the Salt Cell Vent Condenser (SCVC) should also have sufficient capacity to handle the maximum OE boilup. Therefore, it would require simultaneous loss of OEC and/or SCVC condensing capacity for the steam valve failure to cause high benzene in the Process Vessel Vent System (PVVS)

  9. Aerococcus viridans Native Valve Endocarditis

    Directory of Open Access Journals (Sweden)

    Wenwan Zhou

    2013-01-01

    Full Text Available Aerococcus viridans is an infrequent human pathogen and few cases of infective endocarditis have been reported. A case involving a 69-year-old man with colon cancer and hemicolectomy 14 years previously, without recurrence, is reported. A diagnosis of native mitral valve endocarditis was established on the basis of clinical presentation, characteristic echocardiographic findings and pathological specimen examination after urgent valve replacement. A viridans endocarditis appears to be particularly virulent, requiring a surgical approach in four of 10 cases reported and death in one of nine. Given the aggressive nature of A viridans endocarditis and the variable time to diagnosis (a few days to seven months, prompt recognition of symptoms and echocardiography, in addition to blood cultures, should be performed when symptoms persist.

  10. Accidents in nuclear ships

    Energy Technology Data Exchange (ETDEWEB)

    Oelgaard, P L [Risoe National Lab., Roskilde (Denmark); [Technical Univ. of Denmark, Lyngby (Denmark)

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10{sup -3} per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au).

  11. Accidents in nuclear ships

    International Nuclear Information System (INIS)

    Oelgaard, P.L.

    1996-12-01

    This report starts with a discussion of the types of nuclear vessels accidents, in particular accidents which involve the nuclear propulsion systems. Next available information on 61 reported nuclear ship events in considered. Of these 6 deals with U.S. ships, 54 with USSR ships and 1 with a French ship. The ships are in almost all cases nuclear submarines. Only events that involve the sinking of vessels, the nuclear propulsion plants, radiation exposures, fires/explosions, sea-water leaks into the submarines and sinking of vessels are considered. For each event a summary of available information is presented, and comments are added. In some cases the available information is not credible, and these events are neglected. This reduces the number of events to 5 U.S. events, 35 USSR/Russian events and 1 French event. A comparison is made between the reported Soviet accidents and information available on dumped and damaged Soviet naval reactors. It seems possible to obtain good correlation between the two types of events. An analysis is made of the accident and estimates are made of the accident probabilities which are found to be of the order of 10 -3 per ship reactor years. It if finally pointed out that the consequences of nuclear ship accidents are fairly local and does in no way not approach the magnitude of the Chernobyl accident. It is emphasized that some of the information on which this report is based, may not be correct. Consequently some of the results of the assessments made may not be correct. (au)

  12. Bistable diverter valve in microfluidics

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav; Bandulasena, H.C.H.

    2011-01-01

    Roč. 50, č. 5 (2011), s. 1225-1233 ISSN 0723-4864 R&D Projects: GA ČR GA101/07/1499; GA AV ČR IAA200760705 Institutional research plan: CEZ:AV0Z20760514 Keywords : fluidics * bistable diverter valves * pressure-driven microfluidics Subject RIV: BK - Fluid Dynamics Impact factor: 1.735, year: 2011 http://www.springerlink.com/content/x4907p1908151522/

  13. Thermohydraulic accident behavior of reactors

    International Nuclear Information System (INIS)

    Horche, W.; Kirmse, R.; Reichenbach, D.; Weber, J.P.

    1992-01-01

    GRS, on behalf of the German Federal Ministry for the Environment, conducted an assessment of the technical safety of the Greifswald nuclear generating units of the Soviet WWER-440/W-230 and W-213 reactor lines, respectively. The evaluation of existing accident analyses and the execution of some first calculations by GRS added to the know-how of GRS. This is reflected in the increased participation by GRS in international expert bodies investigating safety problems of WWER-440 plants. The contributions made towards international WWER projects within the framework of IAEA missions or as a result of bilateral consultations strengthen international partnership in the field of reactor safety in Central and Eastern Europe. (orig.) [de

  14. Control valve friction operational experience at Darlington NGD

    International Nuclear Information System (INIS)

    Speer, B.

    1995-01-01

    Proper installation of valve packing is an important part of ensuring that control valves operate as intended. Darlington NGD has developed a Valve Packing Program. This program combined with valve diagnostics has enabled the station to ensure that the operability of control valves is maintained after repacking. This paper outlines the process that is used for this. (author)

  15. Transcatheter aortic valve prosthesis surgically replaced 4 months after implantation

    DEFF Research Database (Denmark)

    Thyregod, Hans Gustav; Lund, Jens Teglgaard; Engstrøm, Thomas

    2010-01-01

    Transcatheter aortic valve implantation is a new and rapidly evolving treatment option for high-risk surgical patients with degenerative aortic valve stenosis. Long-term results with these new valve prostheses are lacking, and potential valve dysfunction and failure would require valve replacemen...

  16. Pivot design in bileaflet valves.

    Science.gov (United States)

    Vallana, F; Rinaldi, S; Galletti, P M; Nguyen, A; Piwnica, A

    1992-01-01

    The design criteria leading to the development of a new bileaflet valve (Sorin Bicarbon) were derived from the analysis of functional requirements, the performance of existing prostheses, and the availability of an advanced carbon coating technology (Carbofilm). The hinge is the critical element affecting fluid dynamics, durability, and thrombus formation in bileaflet valves. A comparative study of three existing models led to a new hinge design that was based on coupling two spheric surfaces with different radii of curvature (leaflet pivot and hinge recess) and obtained by electroerosion into a Carbofilm-coated metallic housing. In this valve, the point of contact moves continuously by rolling, not sliding. This minimizes friction and wear and allows uninterrupted washing of the blood exposed surfaces even during diastole (a finding established in patients using transesophageal echocardiography). Tricuspid implantation without anticoagulation in 33 sheep did not lead to thrombotic events (follow-up, 40-400 days). In the first 36 clinical implants observed for 15 months (mitral position, size 29; two unrelated deaths), the mean diastolic gradient by echo Doppler was 4 +/- 1.25 mmHg; the functional area was 3.2 +/- 0.6 cm2. No leaflet fracture and no thrombotic or embolic complications were observed clinically using a standard anticoagulant regimen.

  17. The Fukushima Accident: A Station Blackout and the Consequences

    International Nuclear Information System (INIS)

    Schäfer, F.; Tusheva, P.; Kliem, S.

    2012-01-01

    Lessons learned from Fukushima: • Underestimation of the role of the natural hazards • Insufficient protection of the emergency power and service water systems • Protection of fuel assembly storage pools insufficient • Safety review for Station Blackout and seismic evaluation needed • Diverse power supply systems, diverse sources for water delivery • Role of passive safety systems, they must work in a real passive manner and without electricity to open valves • Backup systems for reactor parameters monitoring • Revision of Severe Accident Management Guidelines and countermeasures for specific “rare” events • Early/late phase operators’ actions / Effectiveness of the operators’ actions

  18. School accidents in Austria.

    Science.gov (United States)

    Schalamon, Johannes; Eberl, Robert; Ainoedhofer, Herwig; Singer, Georg; Spitzer, Peter; Mayr, Johannes; Schober, Peter H; Hoellwarth, Michael E

    2007-09-01

    The aim of this study was to obtain information about the mechanisms and types of injuries in school in Austria. Children between 0 and 18 years of age presenting with injuries at the trauma outpatient in the Department of Pediatric Surgery in Graz and six participating hospitals in Austria were evaluated over a 2-year prospective survey. A total of 28,983 pediatric trauma cases were registered. Personal data, site of the accident, circumstances and mechanisms of accident and the related diagnosis were evaluated. At the Department of Pediatric Surgery in Graz 21,582 questionnaires were completed, out of which 2,148 children had school accidents (10%). The remaining 7,401 questionnaires from peripheral hospitals included 890 school accidents (12%). The male/female ratio was 3:2. In general, sport injuries were a predominant cause of severe trauma (42% severe injuries), compared with other activities in and outside of the school building (26% severe injuries). Injuries during ball-sports contributed to 44% of severe injuries. The upper extremity was most frequently injured (34%), followed by lower extremity (32%), head and neck area (26%) and injuries to thorax and abdomen (8%). Half of all school related injuries occur in children between 10 and 13 years of age. There are typical gender related mechanisms of accident: Boys get frequently injured during soccer, violence, and collisions in and outside of the school building and during craft work. Girls have the highest risk of injuries at ball sports other than soccer.

  19. Force measuring valve assemblies, systems including such valve assemblies and related methods

    Science.gov (United States)

    DeWall, Kevin George [Pocatello, ID; Garcia, Humberto Enrique [Idaho Falls, ID; McKellar, Michael George [Idaho Falls, ID

    2012-04-17

    Methods of evaluating a fluid condition may include stroking a valve member and measuring a force acting on the valve member during the stroke. Methods of evaluating a fluid condition may include measuring a force acting on a valve member in the presence of fluid flow over a period of time and evaluating at least one of the frequency of changes in the measured force over the period of time and the magnitude of the changes in the measured force over the period of time to identify the presence of an anomaly in a fluid flow and, optionally, its estimated location. Methods of evaluating a valve condition may include directing a fluid flow through a valve while stroking a valve member, measuring a force acting on the valve member during the stroke, and comparing the measured force to a reference force. Valve assemblies and related systems are also disclosed.

  20. Design and performance characteristic analysis of servo valve-type water hydraulic poppet valve

    International Nuclear Information System (INIS)

    Park, Sung Hwan

    2009-01-01

    For water hydraulic system control, the flow or pressure control using high-speed solenoid valve controlled by PWM control method could be a good solution for prevention of internal leakage. However, since the PWM control of on-off valves cause extensive flow and pressure fluctuation, it is difficult to control the water hydraulic actuators precisely. In this study, the servo valve-type water hydraulic valve using proportional poppet as the main valve is designed and the performance characteristics of the servo valve-type water hydraulic valve are analyzed. Furthermore, it is demonstrated through experiments that a decline in control chamber pressure that follows the change of pilot flow is caused by the occurrence of cavitation around the proportional poppet, and that fundamental characteristics of the developed valve remain unaffected by the occurrence of cavitation

  1. 3D Printed Multimaterial Microfluidic Valve.

    Directory of Open Access Journals (Sweden)

    Steven J Keating

    Full Text Available We present a novel 3D printed multimaterial microfluidic proportional valve. The microfluidic valve is a fundamental primitive that enables the development of programmable, automated devices for controlling fluids in a precise manner. We discuss valve characterization results, as well as exploratory design variations in channel width, membrane thickness, and membrane stiffness. Compared to previous single material 3D printed valves that are stiff, these printed valves constrain fluidic deformation spatially, through combinations of stiff and flexible materials, to enable intricate geometries in an actuated, functionally graded device. Research presented marks a shift towards 3D printing multi-property programmable fluidic devices in a single step, in which integrated multimaterial valves can be used to control complex fluidic reactions for a variety of applications, including DNA assembly and analysis, continuous sampling and sensing, and soft robotics.

  2. The radiology of prosthetic heart valves

    International Nuclear Information System (INIS)

    Steiner, R.M.; Flicker, S.

    1985-01-01

    The development of prosthetic heart valves in the late 1950s ushered in a new era in the treatment of heart disease. The radiologist has an important role to play preoperatively in the diagnosis of valvular heart disease. Radiology is valuable in identification of the implanted prosthetic valve and recognition of complications associated with valve implantation. Radiologists must be familiar with the imaging techniques best suited to evaluate the function of the valve prosthesis in question. In this chapter the authors discuss the radiographic approach to the evaluation of the status of patients for valve replacement and the imaging problems peculiar to the types of valves in current use. The relative value of plain-film radiography, fluoroscopy, videorecording and cinerecording, and aortography is addressed, as well as the potential value of magnetic resonance imaging and subsecond dynamic computed tomography

  3. NPP Krsko containment environmental conditions during postulated accident

    International Nuclear Information System (INIS)

    Kozaric, M.; Cavlina, N.; Spalj, S.

    1989-01-01

    This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

  4. Evaluation of carbons exposed to the Three Mile Island accident

    International Nuclear Information System (INIS)

    Deitz, V.R.; Romans, J.B.; Bellamy, R.R.

    1981-01-01

    One of the lines of defense that served to mitigate the radiological effects of the accident at Three Mile Island was the activated carbon installed in ventilation air flows. Filters in the Auxiliary and Fuel Handling Buildings of Unit 2 adsorbed tens to hundreds of curies of iodine-131, preventing the release to the environment. The carbon exposed to the accident has been replaced and the spent carbon has been analyzed in the laboratory. Independent analyses were performed for the two filter trains in both the Auxiliary and Fuel Handling Buildings, replaced at various times after the accident. The results of these analyses are compared to new (unexposed) carbons

  5. Feedback from practical experience with large sodium fire accidents

    International Nuclear Information System (INIS)

    Luster, V.P.; Freudenstein, K.F.

    1996-01-01

    The paper reviews the important feedback from the practical experience from two large sodium fires; the first at ALMERIA in Spain and the second in the Na laboratories at Bensberg, Germany. One of the most important sodium fire accidents was the ALMERIA spray fire accident. The origin of this accident was the repair of a valve when about 14 t of sodium was spilled in the plant room over a period of 1/2 hour. The event has been reported (IAEA/IWGFR meeting in 1988) and this presentation gives a short review of important feedback. The Almeria accident was one of the reasons that from that time spray fires had to be taken into account in the safety analyses of nuclear power plants. Due to the fact that spray fire codes were not available in a sufficiently validated state, safety analyses were provisionally based on the feedback from sodium fire tests and also from the Almeria accident itself. The behaviour of spray fires showed that severe destruction, up to melting of metallic structures may occur, but even with a large spray fire is limited roughly within the spray fire zone itself. This could be subsequently be predicted by codes like NABRAND in Germany and FEUMIX in France. Almeria accident has accelerated R and D and code development with respect to spray fires. As example for a code validation some figures are given for the NABRAND code. Another large sodium fire accident happened in 1992 in the test facility at Bensberg in Germany (ILONA). This accident occurred during preheating of a sodium filled vessel which was provisionally installed in the basement of the ILONA test facility at Bensberg. Due to failure of a pressure relief valve the pressure in the vessel increased. As a consequence the plug in a dip tube for draining the vessel failed and about 4,5 t of sodium leaked slowly from the vessel. The plant room was not cladded with steel liners or collecting pans (it was not designed for permanent sodium plant operation). So leaking sodium came directly in

  6. Analysis of hot leg natural circulation under station blackout severe accident

    International Nuclear Information System (INIS)

    Deng Jian; Cao Xuewu

    2007-01-01

    Under severe accidents, natural circulation flows are important to influence the accident progression and result in a pressurized water reactor (PWR). In a station blackout accident with no recovery of steam generator (SG) auxiliary feedwater (TMLB' severe accident scenario), the hot leg countercurrent natural circulation flow is analyzed by using a severe-accident code, to better understand its potential impacts on the creep-rupture timing among the surge line, the hot leg; and SG tubes. The results show that the natural circulation may delay the failure time of the hot leg. The recirculation ratio and the hot mixing factor are also calculated and discussed. (authors)

  7. Trans-apical aortic valve implantation in a patient with stentless valve degeneration.

    Science.gov (United States)

    Kapetanakis, Emmanouil I; MacCarthy, Philip; Monaghan, Mark; Wendler, Olaf

    2011-06-01

    Trans-apical valve-in-valve trans-catheter aortic valve implantation (TAVI) has successfully been performed in selected, high-risk patients, who suffered prosthetic degeneration after aortic valve replacement using stented xenografts. We report the case of a 79-year-old male patient who underwent one of the first successful TAVIs in a failing stentless bioprosthesis. Copyright © 2010 European Association for Cardio-Thoracic Surgery. Published by Elsevier B.V. All rights reserved.

  8. Optimal valve location in long oil pipelines

    OpenAIRE

    Grigoriev, A.; Grigorieva, N.V.

    2007-01-01

    We address the valve location problem, one of the basic problems in design of long oil pipelines. Whenever a pipeline is depressurized, the shutoff valves block the oil flow and seal the damaged part of the pipeline. Thus, the quantity of oil possibly contaminating the area around the pipeline is determined by the volume of the damaged section of the pipeline between two consecutive valves. Then, ecologic damage can be quantified by the amount of leaked oil and the environmental characteristi...

  9. Bistable (latching) solenoid actuated propellant isolation valve

    Science.gov (United States)

    Wichmann, H.; Deboi, H. H.

    1979-01-01

    The design, fabrication, assembly and test of a development configuration bistable (latching) solenoid actuated propellant isolation valve suitable for the control hydrazine and liquid fluorine to an 800 pound thrust rocket engine is described. The valve features a balanced poppet, utilizing metal bellows, a hard poppet/seat interface and a flexure support system for the internal moving components. This support system eliminates sliding surfaces, thereby rendering the valve free of self generated particles.

  10. Improved valve and dash-pot assembly

    Science.gov (United States)

    Chang, S.C.

    1985-04-23

    A dash-pot valve comprises a cylinder submerged in the fluid of a housing and have a piston attached to a plunger projecting into the path of closing movement of a pivotal valve member. A vortex chamber in said cylinder is provided with targentially directed inlets to generate vortex flow upon retraction of said plunger and effect increasing resistance against said piston to progressively retard the closing rate of said valve member toward its seat.

  11. Valve and dash-pot assembly

    Science.gov (United States)

    Chang, Shih-Chih

    1986-01-01

    A dash-pot valve comprising a cylinder submerged in the fluid of a housing and having a piston attached to a plunger projecting into the path of closing movement of a pivotal valve member. A vortex chamber in said cylinder is provided with tangentially directed inlets to generate vortex flow upon retraction of said plunger and effect increasing resistance against said piston to progressively retard the closing rate of said valve member toward its seat.

  12. Radiation accidents and dosimetry

    International Nuclear Information System (INIS)

    Sagstuen, E.; Theisen, H.; Henriksten, T.

    1982-12-01

    On September 2nd 1982 one of the employees of the gamma-irradiation facility at Institute for Energy Technology, Kjeller, Norway entered the irradiation cell with a 65.7 kCi *sp60*Co- source in unshielded position. The victim received an unknown radiation dose and died after 13 days. Using electron spin resonance spectroscopy, the radiation dose in this accident was subsequently determined based on the production of longlived free radicals in nitroglycerol tablets borne by the operator during the accident. He used nitroglycerol for heart problems and free radical are easily formed and trapped in sugar which is the main component of the tablets. Calibration experiments were carried out and the dose given to the tablets during the accident was determined to 37.2 +- 0.5 Gy. The general use of free radicals for dose determinations is discussed. (Auth.)

  13. Big nuclear accidents

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    Much of the debate on the safety of nuclear power focuses on the large number of fatalities that could, in theory, be caused by extremely unlikely but imaginable reactor accidents. This, along with the nuclear industry's inappropriate use of vocabulary during public debate, has given the general public a distorted impression of the safety of nuclear power. The way in which the probability and consequences of big nuclear accidents have been presented in the past is reviewed and recommendations for the future are made including the presentation of the long-term consequences of such accidents in terms of 'reduction in life expectancy', 'increased chance of fatal cancer' and the equivalent pattern of compulsory cigarette smoking. (author)

  14. Care of radiation accidents

    International Nuclear Information System (INIS)

    Renz, K.

    1983-01-01

    The small probability of a serious radiation accident happening dispenses neither the plants where radiation exposure occurs nor the employers' liability insurance associations from their obligation to make provision for such cases. On the other hand, the efforts involved in such preventive measures must be kept within reasonable limits. As a result of these considerations a concept for taking care of radiation accidents was developed that is based on already existing institutions. The most attention was demanded by questions of organization, logistics, communication and information. The syndrome appearing after acute whole-body irradiation is known. This syndrome in its different stages and the relative therapeutic measures form the basis for the organization of the care of radiation accidents. (orig./MG) [de

  15. Exploiting the bead-injection approach in the integrated sequential injection Lab-on-Valve format using hydrophobic packing materials for on-line matrix removal and preconcentration of trace levels of cadmium in environmental and biological samples via formation of non-charged chelates prior

    DEFF Research Database (Denmark)

    Miró, Manuel; Jonczyk, Sylwia; Wang, Jianhua

    2003-01-01

    The concept of renewable microcolumns within the conduits of an automated single injection lab-on-valve system was exploited in a sorption/elution fashion using sorbents of hydrophobic nature. The scheme's practical applicability was demonstrated for the electrothermal atomic absorption spectrome......The concept of renewable microcolumns within the conduits of an automated single injection lab-on-valve system was exploited in a sorption/elution fashion using sorbents of hydrophobic nature. The scheme's practical applicability was demonstrated for the electrothermal atomic absorption...

  16. Review of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Connelly, J.W.; Storr, G.J.

    1989-01-01

    Two types of severe reactor accidents - loss of coolant or coolant flow and transient overpower (TOP) accidents - are described and compared. Accidents in research reactors are discussed. The 1961 SL1 accident in the US is used as an illustration as it incorporates the three features usually combined in a severe accident - a design flaw or flaws in the system, a circumvention of safety circuits or procedures, and gross operator error. The SL1 reactor, the reactivity accident and the following fuel-coolant interaction and steam explosion are reviewed. 3 figs

  17. Double-reed exhaust valve engine

    Science.gov (United States)

    Bennett, Charles L.

    2015-06-30

    An engine based on a reciprocating piston engine that extracts work from pressurized working fluid. The engine includes a double reed outlet valve for controlling the flow of low-pressure working fluid out of the engine. The double reed provides a stronger force resisting closure of the outlet valve than the force tending to open the outlet valve. The double reed valve enables engine operation at relatively higher torque and lower efficiency at low speed, with lower torque, but higher efficiency at high speed.

  18. Fast Flux Test Facility primary sodium valves

    International Nuclear Information System (INIS)

    Rabe, G.B.; Ezra, B.C.

    1977-01-01

    The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

  19. Advantages of butterfly valves for power plants

    International Nuclear Information System (INIS)

    Lapadat, J.T.

    1977-01-01

    Butterfly valves are increasingly used in nuclear power plants. They are used in CANDU reactors for class 2 and 3 service, to provide emergency and tight shutoff valves for all inlets and outlets of heat exchangers and all calandria penetrations. Guidelines for meeting nuclear power plant valve specifications are set out in ASME Section 3, Nuclear Power Plant Components. Some details of materials of construction, type of actuator, etc., for various classes of nuclear service are tabulated in the present article. The 'fishtail' butterfly valve is an improved design with reduced drag, as is illustrated and explained. (N.D.H.)

  20. Additively Manufactured Main Fuel Valve Housing

    Science.gov (United States)

    Eddleman, David; Richard, Jim

    2015-01-01

    Selective Laser Melting (SLM) was utilized to fabricate a liquid hydrogen valve housing typical of those found in rocket engines and main propulsion systems. The SLM process allowed for a valve geometry that would be difficult, if not impossible to fabricate by traditional means. Several valve bodies were built by different SLM suppliers and assembled with valve internals. The assemblies were then tested with liquid nitrogen and operated as desired. One unit was also burst tested and sectioned for materials analysis. The design, test results, and planned testing are presented herein.