The safety relief valve handbook design and use of process safety valves to ASME and International codes and standards

CERN Document Server

Hellemans, Marc

2009-01-01

The Safety Valve Handbook is a professional reference for design, process, instrumentation, plant and maintenance engineers who work with fluid flow and transportation systems in the process industries, which covers the chemical, oil and gas, water, paper and pulp, food and bio products and energy sectors. It meets the need of engineers who have responsibilities for specifying, installing, inspecting or maintaining safety valves and flow control systems. It will also be an important reference for process safety and loss prevention engineers, environmental engineers, and plant and process designers who need to understand the operation of safety valves in a wider equipment or plant design context. . No other publication is dedicated to safety valves or to the extensive codes and standards that govern their installation and use. A single source means users save time in searching for specific information about safety valves. . The Safety Valve Handbook contains all of the vital technical and standards informat...

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

Guidelines for confirmatory inplant tests of safety-relief valve discharges for BWR plants

International Nuclear Information System (INIS)

Su, T.M.

1981-05-01

Inplant tests of safety/relief valve (SRV) discharges may be required to confirm generically established specifications for SRV loads and the maximum suppression pool temperature, and to evaluate possible effects of plant-unique parameters. These tests are required in those plants which have features that differ substantially from those previously tested. Guidelines for formulating appropriate test matrices, establishing test procedures, selecting necessary instrumentation, and reporting the test results are provided in this report. Guidelines to determine if inplant tests are required on the basis of the plant unique parameters are also included in the report

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.

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

Prediction of critical flow rates through power-operated relief valves

International Nuclear Information System (INIS)

Abdollahian, D.; Singh, A.

1983-01-01

Existing single-phase and two-phase critical flow models are used to predict the flow rates through the power-operated relief valves tested in the EPRI Safety and Relief Valve test program. For liquid upstream conditions, Homogeneous Equilibrium Model, Moody, Henry-Fauske and Burnell two-phase critical flow models are used for comparison with data. Under steam upstream conditions, the flow rates are predicted either by the single-phase isentropic equations or the Homogeneous Equilibrium Model, depending on the thermodynamic condition of the fluid at the choking plane. The results of the comparisons are used to specify discharge coefficients for different valves under steam and liquid upstream conditions and evaluate the existing approximate critical flow relations for a wide range of subcooled water and steam conditions

An update to inplace testing of safety/relief valves utilizing lift assist technology

International Nuclear Information System (INIS)

Heorman, K.R.

1992-01-01

Inplace testing of safety and relief valves with lift-assist devices has received mixed reviews from nuclear power plant testing personnel. While many plants use the technology, most limit its use to testing main steam safety valves (even though both OM-1-1981 and PTC 25.3-1976 allow its use for several different service applications). Test coordinator concerns regarding the technology range from lift set point accuracy and repeatability to the quality of the test result output. In addition, OM-1-1981 and PTC 25.3-1976 differ in their approach to the technology. The reasons for the differences between PTC 25.3-1976 and OM-1-1981 are discussed along with additional considerations applicable to the use of the technology in testing liquid service valves. This paper shows that lift assist technology is capable of determining lift set points within the accuracy requirements of OM-1 and PTC 25.3. It also demonstrates that the technology should not be limited to compressible service systems. Also, improvements in test repeatability and output quality are discussed as a function of the assist device design used and valve characteristics. Lift assist testing is often preferred over inplace testing that uses direct system pressure. It is often more cost efficient than bench testing because it does not require removal of critical systems from service and transportation of components. Also, duplicating system temperatures and other environmental factors is not an issue during inplace testing. Valve testing that once required an outage and maintenance period can now be conducted prior to such periods. This approach minimizes the possibility of failures becoming critical path limiting items

State-of-the-Art Report on Pilot Operated Safety Relief Valve

International Nuclear Information System (INIS)

Lee, Jun; Yoon, Ju Hyeon

2003-12-01

The pilot operated safety relief valve(POSRV) is a overpressure protection device. But the POSRV is additionally designed to be functioned as a fast depressurization device differently from the pressurizer safety valve(PSV). So, the POSRV assembly takes on the complicated configuration differently from the PSV, and also its configuration and component parts are much different according to its manufacturers. The objective of the study is to investigate the state of the art for POSRV. We want that the findings of the study are put to practical use as a selection guide to the utilities or system designers which want to purchase the POSRV. In this study, we selected the SEBIM and CCI made goods as the subject of investigation, and have investigated the technical characteristics and operating principles of the goods. However, in this report, the comparative contents with other company which were issued by one company were not handled. Also the advantages or disadvantages between two companies were not handled. Only the technical characteristics and operating principles with each company were described. For reference, the main considerations to select the POSRV' type are follows. - the characteristics of the system that the POSRV is to be installed - the fitness of the POSRV for the system - the requirements related to the function and operation of the POSRV - the vender's opinion

Soil-structure interaction for transient loads due to safety relief valve discharges

International Nuclear Information System (INIS)

Tseng, W.S.; Tsai, N.C.

1978-01-01

Dynamic responses of BWR Mark II containment structures subjected to axisymmetric transient pressure loadings due to simultaneous safety relief valve discharges were investigated using finite element analysis, including the soil-structure interaction effect. To properly consider the soil-structure interaction effect, a simplified lumped parameter foundation model and axisymmetric finite element foundation model with viscous boundary impedance are used. Analytical results are presented to demonstrate the effectiveness of the simplified foundation model and to exhibit the dynamic response behavior of the structure as the transient loading frequency and the foundation rigidity vary. The impact of the dynamic structural response due to this type of loading on the equipment design is also discussed. (Auth.)

New safety valve addresses environmental concerns

International Nuclear Information System (INIS)

Taylor, J.; Austin, R.

1992-01-01

This paper reports that Conoco Pipeline is using a unique relief valve to reduce costs while improving environmental protection at its facilities. Conoco Pipeline Co. Inc. began testing new relief valves in 1987 to present over-pressuring its pipelines while enhancing the safety, environmental integrity and profitability of its pipelines. Conoco worked jointly with Rupture Pin Technology Inc., Oklahoma City, to seek a solution to a series of safety, environmental, and operational risks in the transportation of crude oil and refined products through pipelines. Several of the identified problems were traced to a single equipment source: the reliability of rupture discs used at pipeline stations to relieve pressure by diverting flow to tanks during over-pressure conditions. Conoco's corporate safety and environmental policies requires solving problems that deal with exposure to hydrocarbon vapors, chemical spills or the atmospheric release of fugitive emissions, such as during rupture disc maintenance. The company had used rupture pin valves as vent relief devices in conjunction with development by Rick Austin of inert gas methods to protect the inner casing wall and outer carrier pipeline wall in pipeline road crossings. The design relies on rupture pin valves set at 5 psi to isolate vent openings from the atmosphere prior to purging the annular space between the pipeline and casing with inert gas to prevent corrosion. Speciality Pipeline Inspection and Engineering Inc., Houston, is licensed to distribute the equipment for the new cased-crossing procedure

Dynamic load in suppression pool during BWR main steam safety relief valve actuation

International Nuclear Information System (INIS)

Tsukada, Hiroshi; Yamaguchi, Hirokatsu; Morita, Terumichi

1979-01-01

BWRs are so designed that the exhaust steam from main steam safety relief valves is led to pressure suppression pools, and the steam is condensed in pool water, but at this time, dynamic load seems to arise in the pool water. In Tokai No. 2 Power Station, a Mark-2 containment vessel was adopted to improve the reliability as much as possible and to obtain the design with margin. In this report, the result of actual machine test in Tokai No. 2 Power Station and the method of reducing the load are described. When a relief valve works, the discharge of water in exhaust pipes into a suppression pool, the exhaust of air in exhaust pipes and repeated expansion and contraction of bubbles in pool water, and the exhaust of steam and condensation occur. As for the construction of the suppression pool in Tokai No. 2 Power Station, cross-shaped quencher and the structure with jet deflector were installed. The test plan and the test result with an actual machine are reported. The soundness of the Mark-2 containment vessel and the structures in the pool was proved. The differential pressure acting on the structures was negligibly small. The measured pulsating pressure was in the range from 0.84 to -0.39 kg/cm 2 . (Kako, I.)

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

Numerical and experimental investigation on the performance of safety valves operating with different gases

International Nuclear Information System (INIS)

Dossena, V.; Marinoni, F.; Bassi, F.; Franchina, N.; Savini, M.

2013-01-01

A detailed analysis of the effect related to the expansion of different gases throughout safety relief valves is carried out both numerically and experimentally. The considered gases are air, argon and ethylene, representative of a wide range of specific heat ratios. A first experimental campaign performed in air and argon on a safety relief valve characterized by connection 1/2″ × 1″ and orifice designation D (diameter 10 mm) according to API 526 showed significant reduction both in disc lift and in exhausted mass flow rate, at the nominal overpressure, when operating with argon. In order to gain a deeper insight into the physics involved and to evaluate the valve behavior with other gases, an extensive numerical testing has been performed by means of an accurate CFD code based on discontinuous Galerkin formulation. Numerical results are at first validated against measurements obtained in air on a 2″ J 3″ safety relief valve proving a remarkable accuracy of the computational method. Then the validated solver is applied on the same computational grid using argon and ethylene as working fluids. The three gases are considered as thermally perfect gases. A critical discussion based on the numerical results allows to clarify the fluid dynamic and physical reasons causing the observed trends both in the opening force and in the discharge coefficient. The main conclusion is that particular care must be taken when a safety valve operates with a fluid characterized by a specific heat ratio greater than the one of the gas used during type testing. -- Highlights: ► Effects of different gases on the discharge capacity and operational characteristics on safety relief valves. ► Influence of different specific heat ratio on safety relief valves discharge coefficient. ► Skilful application of Discontinuous Galerkin CFD solver to safety valves performances prediction

Technical evaluation report TMI action - NUREG-0737 (II.D.1) relief and safety valve testing for Clinton Power Station Unit 1. (Docket No. 50-461)

International Nuclear Information System (INIS)

Burr, T.K.; Magleby, H.L.

1985-05-01

Light water reactors operators have experienced a number of occurrences of improper performance by safety and relief valves installed in their primary coolant systems. Because of this, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) recommended that programs be developed and completed which would reevaluate the performance capabilities of BWR safety and relief valves. This report has examined the response of the Licensee for the Clinton Power Station, Unit 1 to the requirements of NUREG-0578 and subsequently NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15 and 30 of Appendix A to 10 CFR-50 have been met

Analysis and qualification of steam generator relief valves (BRU-A)

International Nuclear Information System (INIS)

Lathuile, C.; Serre, J. L.

1997-01-01

This paper presents a general overview of improvements foreseen in the frame of Safety Measures S01 and S10 in order to prevent and mitigate consequences of a large primary to secondary leakage. Among these improvements, a more detailed description of methodology and results relative to Steam Generator Relief Valves (BRU-A) qualification tests is presented. (author)

Method of accounting for code safety valve setpoint drift in safety analyses

International Nuclear Information System (INIS)

Rousseau, K.R.; Bergeron, P.A.

1989-01-01

In performing the safety analyses for transients that result in a challenge to the reactor coolant system (RCS) pressure boundary, the general acceptance criterion is that the peak RCS pressure not exceed the American Society of Mechanical Engineers limit of 110% of the design pressure. Without crediting non-safety-grade pressure mitigating systems, protection from this limit is mainly provided by the primary and secondary code safety valves. In theory, the combination of relief capacity and setpoints for these valves is designed to provide this protection. Generally, banks of valves are set at varying setpoints staggered by 15- to 20-psid increments to minimize the number of valves that would open by an overpressure challenge. In practice, however, when these valves are removed and tested (typically during a refueling outage), setpoints are sometimes found to have drifted by >50 psid. This drift should be accounted for during the performance of the safety analysis. This paper describes analyses performed by Yankee Atomic Electric Company (YAEC) to account for setpoint drift in safety valves from testing. The results of these analyses are used to define safety valve operability or acceptance criteria

A computational method to predict fluid-structure interaction of pressure relief valves

Energy Technology Data Exchange (ETDEWEB)

Kang, S. K.; Lee, D. H.; Park, S. K.; Hong, S. R. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

2004-07-01

An effective CFD (Computational fluid dynamics) method to predict important performance parameters, such as blowdown and chattering, for pressure relief valves in NPPs is provided in the present study. To calculate the valve motion, 6DOF (six degree of freedom) model is used. A chimera overset grid method is utilized to this study for the elimination of grid remeshing problem, when the disk moves. Further, CFD-Fastran which is developed by CFD-RC for compressible flow analysis is applied to an 1' safety valve. The prediction results ensure the applicability of the presented method in this study.

Numerical investigation on cavitation in pressure relief valve for coal liquefaction

International Nuclear Information System (INIS)

Ou, G F; Li, W Z; Xiao, D H; Zheng, Z J; Dou, H S; Wang, C

2015-01-01

The pressure relief valve for regulating the level of the high-pressure separator works under a pressure difference up to 15 MPa in the temperature of 415 °C. Severe cavitation erosion and particle impact lead to the valve disc's mass loss. In this paper, three-dimensional turbulent cavitating flows in the pressure relief valve are numerically simulated to reveal the mechanism of mass loss at valve disc. The RNG k-ε turbulence model and the mixture model with a mass transfer for cavitation are employed to simulate the cavitating flow in the pressure relief valve. The result shows that there is phase change in the pressure relief process and cavitation bubbles would be transported by high-velocity backflow to the head of valve disc. For the local pressure higher than the saturated vapor pressure, the bubbles collapse at the head of disc and cavitation erosion is formed at the head of the disc. By comparing the cases of opening of 40%, 50%, and 60%, backflow velocity and cavitation region in front of the disc decrease with the opening increase. Therefore, during the actual operation, the pressure relief valve should be kept to a relatively large opening

Study on high reliability safety valve for railway vehicle

Science.gov (United States)

Zhang, Xuan; Chen, Ruikun; Zhang, Shixi; Xu, BuDu

2017-09-01

Now, the realization of most of the functions of the railway vehicles rely on compressed air, so the demand for compressed air is growing higher and higher. This safety valve is a protection device for pressure limitation and pressure relief in an air supply system of railway vehicles. I am going to introduce the structure, operating principle, research and development process of the safety valve designed by our company in this document.

Mechanical Designs for Relief Valves for Cryogenic Apparatuses and Installations

CERN Multimedia

CERN. Geneva

2016-01-01

There are also pressure relief valves with warm seat available on which the set pressure is based on an adjustment of forces by permanent magnets. Pressure vessel rules allows also the choice for an active triggered pressure relief valve (Cont...

Impact of valve failures on the safety and reliability of light water nuclear power plants

International Nuclear Information System (INIS)

Riddington, J.W.; Reyer, R.J.

1980-01-01

A study of the causes of, and solutions for, recurrent valve failures has been performed. The frequency and root causes of valve problems were identified from licensee event reports and meetings with utility, NSSS, and valve manufacturer personnel. Three generic problems (stem leakage, seat leakage, and inadequate specification) and four valve specific problems were identified. The four valve specific problems and their principal causes are: (1) BWR pilot operated safety relief valves (pilot valve leakage); (2) spring loaded safety relief valves (water solid and two-phase flow behavior); (3) PWR feedwater regulating valves (trim degradation and packing failures); and (4) air operated solenoid valves (jamming due to foreign matter in service air). The first two valve specific problems are the subject of current industry programs. Programs intended to address stem leakage, seat leakage, timely exchange of valve failure information, testing of valves, and adequate specification, selection, and maintenance of valves will be outlined

Processing techniques for data from the Kuosheng Unit 1 shakedown safety-relief-valve tests

International Nuclear Information System (INIS)

McCauley, E.W.; Rompel, S.L.; Weaver, H.J.; Altenbach, T.J.

1982-08-01

This report describes techniques developed at the Lawrence Livermore National Laobratory, Livermore, CA for processing original data from the Taiwan Power Company's Kuosheng MKIII Unit 1 Safety Relief Valve Shakedown Tests conducted in April/May 1981. The computer codes used, TPSORT, TPPLOT, and TPPSD, form a special evaluation system for treating the data from its original packed binary form to ordered, calibrated ASCII transducer files and then to production of time-history plots, numerical output files, and spectral analyses. Using the data processing techniques described, a convenient means of independently examining and analyzing a unique data base for steam condensation phenomena in the MARKIII wetwell is described. The techniques developed for handling these data are applicable to the treatment of similar, but perhaps differently structured, experiment data sets

Safety/relief valve quencher loads: evaluation for BWR Mark II and III containments

International Nuclear Information System (INIS)

Su, T.M.

1982-10-01

Boiling water reactor (BWR) plants are equipped with safety/relief valves (SRVs) to protect the reactor from overpressurization. Plant operational transients, such as turbine trips, will actuate the SRV. Once the SRV opens, the air column within the partially submerged discharge line is compressed by the high-pressure steam released from the reactor. The compressed air discharged into the suppression pool produces high-pressure bubbles. Oscillatory expansion and contraction of these bubbles create hydrodynamic loads on the containment structures, piping, and equipment inside containment. This report presents the results of the staff's evaluation of SRV loads. The evaluation, however, is limited to the quencher devices used in Mark II and III containments. With respect to Mark I containments, the SRV acceptance criteria are presented in NUREG-0661 issued July 1980. The staff acceptance criteria for SRV loads for Mark II and III containments are presented in this report

Technical evaluation report - TMI action: NUREG-0737 (II.D.1) relief and safety valve testing for Grand Gulf Nuclear Station Unit No. 1 (Docket No. 50-416)

International Nuclear Information System (INIS)

Burr, T.K.; Nalezny, C.L.

1985-09-01

Light water reactors operators have experienced a number of occurrences of improper performance by safety and relief valves installed in their primary coolant systems. Because of this, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) recommended that programs be developed and completed which would reevaluate the performance capabilities of BWR safety and relief valves. This report provides the results of the review of these programs and their results by the NRC and their consultant, EG and G Idaho, Inc. Specifically, this report has examined the response of the Licensee for the Grand Gulf Nuclear Station, Unit 1 to the requirements of NUREG-0578 and subsequently NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15 and 30 of Appendix A to 10 CFR-50 have been met

A Study on the Main Steam Safety Valve Opening Mechanism by Flashing on NPPs

International Nuclear Information System (INIS)

Kim, Bae Joo

2009-01-01

A safety injection event happened by opening of the Main Steam Safety Valve at Kori unit 1 on April 16, 2005. The safety valves were opened at the lower system pressure than the valve opening set point due to rapid system pressure drop by opening of the Power Operated Relief Valve installed at the upstream of the Main Steam System. But the opening mechanism of safety valve at the lower set point pressure was not explained exactly. So, it needs to be understood about the safety valve opening mechanism to prevent a recurrence of this kind of event at a similar system of Nuclear Power Plant. This study is aimed to suggest the hydrodynamic mechanism for the safety valve opening at the lower set point pressure and the possibility of the recurrence at similar system conditions through document reviewing for the related previous studies and Kori unit 1 event

Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

International Nuclear Information System (INIS)

Garbett, K.; Mendler, O.J.; Gardner, G.C.; Garnsey, R.; Young, M.Y.

1987-03-01

In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faults and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated

Fluid-structure interaction dynamic simulation of spring-loaded pressure relief valves under seismic wave

Science.gov (United States)

Lv, Dongwei; Zhang, Jian; Yu, Xinhai

2018-05-01

In this paper, a fluid-structure interaction dynamic simulation method of spring-loaded pressure relief valve was established. The dynamic performances of the fluid regions and the stress and strain of the structure regions were calculated at the same time by accurately setting up the contact pairs between the solid parts and the coupling surfaces between the fluid regions and the structure regions. A two way fluid-structure interaction dynamic simulation of a simplified pressure relief valve model was carried out. The influence of vertical sinusoidal seismic waves on the performance of the pressure relief valve was preliminarily investigated by loading sine waves. Under vertical seismic waves, the pressure relief valve will flutter, and the reseating pressure was affected by the amplitude and frequency of the seismic waves. This simulation method of the pressure relief valve under vertical seismic waves can provide effective means for investigating the seismic performances of the valves, and make up for the shortcomings of the experiment.

RELAP5/MOD3 assessment for calculation of safety and relief valve discharge piping hydrodynamic loads

International Nuclear Information System (INIS)

Stubbe, E.J.; VanHoenacker, L.; Otero, R.

1994-02-01

This report presents an assessment study for the use of the code RELAP 5/MOD3/5M5 in the calculation of transient hydrodynamic loads on safety and relief discharge pipes. Its predecessor, RELAP 5/MOD1, was found adequate for this kind of calculations by EPRI. The hydrodynamic loads are very important for the discharge piping design because of the fast opening of the valves and the presence of liquid in the upstream loop seals. The code results are compared to experimental load measurements performed at the Combustion Engineering Laboratory in Windsor (US). Those measurements were part of the PWR Valve Test Program undertaken by EPRI after the TMI-2 accident. This particular kind of transients challenges the applicability of the following code models: two-phase choked discharge; interphase drag in conditions with large density gradients; heat transfer to metallic structures in fast changing conditions; two-phase flow at abrupt expansions. The code applicability to this kind of transients is investigated. Some sensitivity analyses to different code and model options are performed. Finally, the suitability of the code and some modeling guidelines are discussed

Implementation of an enlarged model of the safety valves and relief in the plant integral model for the code RELAP/SCDAPSIM; Implementacion de un modelo ampliado de las valvulas de seguridad y alivio en el modelo integral de planta para el codigo RELAP/SCDAPSIM

Energy Technology Data Exchange (ETDEWEB)

Amador G, R.; Ortiz V, J.; Castillo D, R. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Hernandez L, E. J. [Instituto Tecnologico de Toluca, Av. Tecnologico s/n, Fracc. La Virgen, 52149 Metepec, Estado de Mexico (Mexico); Galeana R, J. C. [Universidad del Valle de Mexico, Campus Toluca, Av. de Las Palmas 136, 52140 Metepec, Estado de Mexico (Mexico); Gutierrez, V. H., E-mail: rodolfo.amador@inin.gob.mx [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Av. San Pablo 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico)

2013-10-15

The present work refers to the implementation of a new model on the logic of the safety valves and relief in the integral model of the Nuclear Power Plant of Laguna Verde of the thermal-hydraulic compute code RELAP/SCDAPSIM Mod. 3.4. The new model was developed with the compute package SIMULINK-MATLAB and contemplates all the operation options of the safety valves and relief, besides including the availability options of the valves in all the operation ways and of blockage in the ways of relief and low-low. The implementation means the elimination of the old model of the safety valves and to analyze the group of logical variables, of discharge and available control systems to associate them to the model of package SIMULINK-MATLAB. The implementation has been practically transparent and 27 cases corresponding to a turbine discharge were analyzed with the code RELAP/SCDAPSIM Mod. 3.4. The results were satisfactory. (Author)

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

Technical evaluation report TMI action: NUREG-0737 (II.D.1) relief and safety valve testing,. Diablo Canyon Units 1 and 2 (Docket Nos. 50-275, 50-323)

International Nuclear Information System (INIS)

Miller, G.K.; Magleby, H.L.; Nalezny, C.L.

1984-07-01

Light water reactor operators have experienced a number of occurrences of improper performance of safety and relief valves installed in their primary coolant systems. Because of this, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Status Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient and accident conditions. This report provides the results of the review of these programs and their results by the Nuclear Regulatory Commission (NRC) and their consultant, EG and G Idaho, Inc. Specifically, this report has examined the response of the Licensee for Diablo Canyon Units 1 and 2, to the requirements of NUREG-0578 and NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15 and 30 of Appendix A to 10 CRF 50 have been met. 18 refs

Examples, clarifications, and guidance on preparing requests for relief from pump and valve inservice testing requirements

International Nuclear Information System (INIS)

Ransom, C.B.; Hartley, R.S.

1996-02-01

In this report, the Idaho National Engineering Laboratory reviewers discuss related to requests for relief from the American Society of Mechanical Engineers code requirements for inservice testing (IST) of safety-related pumps and valves at commercial nuclear power plants. This report compiles information and examples that may be useful to licensees in developing relief requests submitted to US Nuclear Regulatory Commission (NRC) for their consideration and provides insights and recommendations on related IST issues. The report also gives specific guidance on relief requests acceptable and not acceptable to the NRC and advises licensees in the use of this information for application at their facilities

Root cause evaluation of pressurizer relief valve leakage

International Nuclear Information System (INIS)

Olson, D.E.; Voll, B.J.

1996-01-01

Pressurizer relief valves at two pressurized water reactor units experienced unacceptable leakage during plant heatup. The leakage was suspected to be caused by excessive pipe loads on the valves. This paper describes how monitoring via hard-wired transducers and a digital data acquisition system was used to quantify the pipe loads on the valves, and assist in determining the root cause of the pipe loads and appropriate corrective actions. The selection of the parameters monitored, how the monitoring was accomplished and interpretation of the results is discussed. The corrective actions implemented based on the monitoring results are also discussed

Standard Practice for Installation, Inspection, and Maintenance of Valve-body Pressure-relief Methods for Geothermal and Other High-Temperature Liquid Applications

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This practice covers installation, inspection, and maintenance of valve body cavity pressure relief methods for valves used in geothermal and other high-temperature liquid service. The valve type covered by this practice is a design with an isolated body cavity such that when the valve is in either the open or closed position pressure is trapped in the isolated cavity, and there is no provision to relieve the excess pressure internally. 1.2 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

A conceptual study on large-capacity safety relief valve (SRV) for future BWR plants

International Nuclear Information System (INIS)

Yamada, Katsumi; Tokunaga, Takashi; Iwanaga, Masakazu; Kurosaki, Toshikazu

1999-01-01

This paper presents a conceptual study of Safety Relief Valve (SRV) which has larger flow capacity than that of the conventional one and a new structure. Maintenance work of SRVs is one of the main concerns for next-generation Boiling Water Reactor (BWR) plants whose thermal power is planned to be increased. Because the number of SRVs increases with the thermal power, their maintenance would become critical during periodic inspections. To decrease the maintenance work, reduction of the number by increasing the nominal flow rate per SRV and a new structure suitable for easier treatment have been investigated. From a parameter survey of the initial and maintenance cost, the optimum capacity has been estimated to be between 180 and 200 kg/s. Primarily because the number of SRVs decreases in inversely proportional to the capacity, the total maintenance work decreases. The new structure of SRV, with an internally mounted actuator, decreases the number of the connecting parts and will make the maintenance work easier. A 1/4-scale model of the new SRV has been manufactured and performance tests have been conducted. The test results satisfied the design target, which shows the feasibility of the new structure. (author)

Discharge models through the pressurizer valves

International Nuclear Information System (INIS)

Madeira, A.A.

1985-01-01

A reliable estimate of discharge through the pressurizer relief and safety valves is of concern to adequately predict the behaviour of RCS pressure during transients. It's investigated the discharge models used by the ALMOD code, and to implement alternative models from the available literature, which are recommended for different conditions of flow that shall exist during transients requiring discharge through the relief and safety valves. (Author) [pt

46 CFR 53.05-1 - Safety valve requirements for steam boilers (modifies HG-400 and HG-401).

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Safety valve requirements for steam boilers (modifies HG-400 and HG-401). 53.05-1 Section 53.05-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY... requirements for steam boilers (modifies HG-400 and HG-401). (a) The pressure relief valve requirements and the...

Leakage characterization of a piloted power operated relief valve

International Nuclear Information System (INIS)

Ezekoye, L.I.; Hess, M.D.

1995-01-01

In Westinghouse Pressurized Water Reactors (PWRs), power operated relief valves (PORVs) are used to provide overpressure protection of the Pressurizer. The valves are fail closed globe valves which means that power is required to open the valves and, on loss of power, the valves close. There are two ways to operate the PORVs. The more common way is to directly couple the disc to an actuator via a disc-stem assembly. The type of design is not the object of this paper. The other and less common way of operating a PORV is by piloting the main valve such that the opening or closing of a pilot valve opens and closes the main valve. This is the design of interest. In most plants, the PORVs are installed with a water loop seal while in some plants no water loop seals are used. It is generally accepted that loop seal installation minimizes valve seat leakage. In non-loop seal installation, the valve seat is exposed to steam which increases the potential for seat leakage. This paper describes the results of some tests performed with nitrogen and steam to characterize the leakage potential of a pilot operated PORV. The test results were compared with seat leakage tests of check valves to provide insight on the leakage testing of pilot operated valves and check valves. The paper also compares the test data with leakage estimates using the ASME/ANSI OM Code guidance on valve leakage

46 CFR 53.05-2 - Relief valve requirements for hot water boilers (modifies HG-400.2).

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Relief valve requirements for hot water boilers (modifies HG-400.2). 53.05-2 Section 53.05-2 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY... requirements for hot water boilers (modifies HG-400.2). (a) The relief valve requirements for hot water boilers...

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)

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...

Crygenic performance of a superfluid helium relief valve for the LHC superconducting magnets

International Nuclear Information System (INIS)

Danielsson, H.; Ferlin, G.; Luguet, C.

1996-01-01

The high-field superconducting magnets of the Large Hadron Collider (LHC) project at CERN will operate below 1.9 K in static baths of pressurized helium II. In case of resistive transition (open-quotes quenchclose quotes), the resulting pressure rise in the cryostats must be limited to below their 2 MPa design pressure. This is achieved by discharging helium at high flow-rates into a cold recovery header, normally maintained at 20 K. For this purpose, the authors have designed, built and tested a cryogenic quench relief valve with a nominal diameter of 50 mm and an opening time of below 0.1 s. The valve, which can be opened on an external trigger, also acts as a relief device actuated by the upstream pressure when it exceeds 0.4 MPa. In normal operation, the closed poppet must be helium-tight, for hydraulic and thermal separation of the magnet baths from the recovery header. Following mechanical qualification tests under vacuum, the authors have mounted the relief valve in a dedicated cryogenic measuring bench, in order to perform precision thermal measurements with pressurized helium II

Modelling of coupled self-actuating safety, relief and damped check valve systems with the codes TRAC-PF1 and ROLAST

International Nuclear Information System (INIS)

Neumann, U.; Puzalowski, R.; Grimm, I.

1985-01-01

Numerical valve models for simulation of selfactuating safety valves and damped check valves are introduced for the computer programs TRAC-PF1 and ROLAST. As examples of application post-test calculations and stability analysis are given. (orig.)

Safety valve opening and closing operation monitor

International Nuclear Information System (INIS)

Kodama, Kunio; Takeshima, Ikuo; Takahashi, Kiyokazu.

1981-01-01

Purpose: To enable the detection of the closing of a safety valve when the internal pressure in a BWR type reactor is a value which will close the safety valve, by inputting signals from a pressure detecting device mounted directly at a reactor vessel and a safety valve discharge pressure detecting device to an AND logic circuit. Constitution: A safety valve monitor is formed of a pressure switch mounted at a reactor pressure vessel, a pressure switch mounted at the exhaust pipe of the escape safety valve and a logic circuit and the lide. When the input pressure of the safety valve is raised so that the valve and the pressure switch mounted at the exhaust pipe are operated, an alarm is indicated, and the operation of the pressure switch mounted at a pressure vessel is eliminated. If the safety valve is not reclosed when the vessel pressure is decreased lower than the pressure at which it is to be reclosed after the safety valve is operated, an alarm is generated by the logic circuit since both the pressure switches are operated. (Sekiya, K.)

Investigation Of Adhesion Formation In New Stainless Steel Trim Spring Operated Pressure Relief Valves

Energy Technology Data Exchange (ETDEWEB)

Gross, Robert E. [Savannah River Site (SRS), Aiken, SC (United States); Bukowski, Julia V. [Villanova University, Villanova, PA (United States); Goble, William M. [exida, Sellersville, PA (United States)

2013-04-16

Examination of proof test data for new (not previously installed) stainless steel (SS) trim spring operated pressure relief valves (SOPRV) reveals that adhesions form between the seat and disc in about 46% of all such SOPRV. The forces needed to overcome these adhesions can be sufficiently large to cause the SOPRV to fail its proof test (FPT) prior to installation. Furthermore, a significant percentage of SOPRV which are found to FPT are also found to ''fail to open'' (FTO) meaning they would not relief excess pressure in the event of an overpressure event. The cases where adhesions result in FTO or FPT appear to be confined to SOPRV with diameters < 1 in and set pressures < 150 psig and the FTO are estimated to occur in 0.31% to 2.00% of this subpopulation of SS trim SOPRV. The reliability and safety implications of these finding for end-users who do not perform pre-installation testing of SOPRV are discussed.

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.

Static characteristics of a pilot relief valve; Baransupisuton gata ririfu ben no sei tokusei

Energy Technology Data Exchange (ETDEWEB)

Washio, S.; Yonguang YU; Nakamura, Y. [Okayama Univ. (Japan). School of Engineering

1997-09-25

Among a lot of hydraulic valves, relief valve is the most fundamental and important valve which takes change of pressure control. It is essential to know the working characteristics accurately, in order to predict the performance of . system. So far, the numerical simulation of relief valve has been tried and the mathematical models of individual component characteristics based on the knowledge of hydraulics are proposed. In this report, it was revealed that the static relation among the pressure drop, flow rate and opening area for a constriction can be represented, not by the traditional hydraulic orifice equation which has always been used for the purpose but by a new one including an additional pressure loss proportional to the flow rate and the fluid viscosity and inversely proportional to the square of the opening area. The new characteristic equation has proved to consistently forecast the experimental findings in which the rise in oil temperature results in an increase in the piston displacement, but causes little changes as regards regulated pressure. It has also turned out that contrary to conventional preconception, the fluid force exerted on a poppet is negligible. 6 refs., 14 figs., 1 tab.

SEBIM pilot operated tandems. A new solution for Darlington NGS bleed condenser relief valves

International Nuclear Information System (INIS)

Paetzold, H.; Hera, V.; Schaumburg, G.

1996-01-01

Following incidents at Pickering, Wolsung and Bruce NGS, involving instability of bleed condenser relief valves, Darlington station decided to replace the spring loaded RV's by new pilot operated SEBIM tandem valves. This paper is presenting the approach taken, the design and the testing of the new solution, as well as some of the computer modeling work performed in connection with this project. The SEBIM tandems, following successful testing in France, will be installed in Darlington Unit 2, this spring. The new valves can perform with absence of instability and prevent a LOCA incident due to their design, which includes a protection and a redundant valve in series. (author)

46 CFR 61.05-20 - Boiler safety valves.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Boiler safety valves. 61.05-20 Section 61.05-20 Shipping... INSPECTIONS Tests and Inspections of Boilers § 61.05-20 Boiler safety valves. Each safety valve for a drum, superheater, or reheater of a boiler shall be tested at the interval specified by table 61.05-10. [CGD 95-028...

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

Qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1981-01-01

This Standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. It may also be used to separately qualify actuator components. This Standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

A correlation for safety valve blowdown and ring settings

International Nuclear Information System (INIS)

Singh, A.; Shak, D.

1982-01-01

The blowdown of a spring loaded safety valve is defined as the difference between the pressure at which the valve opens and the pressure at which the valve fully closes under certain fluid flow conditions. Generally, the blowdown is expressed in terms of percentage of the opening pressure. An extensive series of tests carried out in the EPRI/PWR Utilities Valve Test Program has shown that the blowdown of safety valves can in general be strongly dependent upon the valve geometry and other parameters such as ring adjustments, spring stiffness, backpressure etc. In the present study, correlations have been developed using the EPRI safety valve test data to predict the expected blowdown as a function of adjustment ring settings for geometrically similar valves under steam discharge conditions. The correlation is validated against two different size Dresser valves

Potential problems will drillstring safety valves

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-06-01

In the paper titled ``New generation drill string safety valves,`` presented at the IADC Well Control Conference for Europe, Aberdeen, May 22--24, 1996, documented limitations of presently available drillstring safety valves commonly used as kelly valves and stabbing valves were presented, and industry efforts to develop solutions to these problems were described. Authors of the paper are B.A. Tarr and R.A. Sukup (Mobil E and P Technology Center, Dallas), Dr. R. Luy (ITE, Clausthal, Germany), G. Rabby (Hi-Kalibre, Edmonton, Alberta) and J. Mertsch (ITAG, Celle, Germany). In 1995, the Task Group developed a draft of a new spec, and a DSSV testing program was initiated as a joint industry project, with the Gas Research Institute (GRI) as the major sponsor. Two manufacturers, Hi-Kalibre and ITAG, agreed to build new valves for the testing program. Hi-Kalibre, Edmonton, Alberta, supplied an already-commercial twin floating ball valve for November 1995 testing. This product is being used by Tesco in its portable top drive system. ITAG of Germany supplied a radically improved DSSV design, which was tested in December, and was to be retested in May following modifications.

Development of Proportional Pressure Control Valve for Hydraulic Braking Actuator of Automobile ABS

Directory of Open Access Journals (Sweden)

Che-Pin Chen

2018-04-01

Full Text Available This research developed a novel proportional pressure control valve for an automobile hydraulic braking actuator. It also analyzed and simulated solenoid force of the control valves, and the pressure relief capability test of electromagnetic thrust with the proportional valve body. Considering the high controllability and ease of production, the driver of this proportional valve was designed with a small volume and powerful solenoid force to control braking pressure and flow. Since the proportional valve can have closed-loop control, the proportional valve can replace a conventional solenoid valve in current brake actuators. With the proportional valve controlling braking and pressure relief mode, it can narrow the space of hydraulic braking actuator, and precisely control braking force to achieve safety objectives. Finally, the proposed novel proportional pressure control valve of an automobile hydraulic braking actuator was implemented and verified experimentally.

The modification of main steam safety valves in Qinshan phase Ⅱ expansion project

International Nuclear Information System (INIS)

Chen Haiqiao

2012-01-01

The main steam safety valves of NPP steam system are second- class nuclear safety component. It used to limit the pressure of SG secondary side and main steam system via emitting steam into the environment. At present, the main steam safety valves have mechanical valves and assisted power valves. According to the experience of power plants at home and abroad, including Qinshan Phase Ⅱ unit 1/2 experience feedback, Qinshan Phase Ⅱ expansion project made modification on valve type, setting value and valve body. This paper introduce the characteristics of different safety valve types, the modification of main steam safety valves and the modification analysis on safety issues.security and impact on the other systems in Qinshan Phase Ⅱ expansion project. (author)

Analysis of liquid relief valves opening demand during pressure increase abnormal scenarios at Embalse nuclear power plant

International Nuclear Information System (INIS)

Bedrossian, Gustavo C.; Gersberg, Sara

2000-01-01

Two hypothetical scenarios have been analyzed where, after an initiating event, Embalse nuclear power plant primary heat transport system could undergo a pressure increase. These abnormal events are a loss of feedwater to the steam generators and a loss of Class IV power supply with Class III restoration. This analysis focuses on primary system liquid relief valves action, specially on their opening demand. Calculation results show that even when these valves are expected to open during the transient, primary system maximum allowable pressure would not be exceeded if they failed to open. System response was also studied in case that one of these relief valves did not close once primary system pressure decreases. For the scenario of loss of feedwater to steam generators, if the degasser-condenser could not be bottled-up, Emergency Cooling Injection conditions would be reached due to a continuos loss of coolant. In case of loss of Class IV -and assuming degasser-condenser bottling-up as service water would not be available- it was observed that primary system should remain pressurized, and with core cooled by thermo siphoning mechanism. (author)

VALIDATION OF SPRING OPERATED PRESSURE RELIEF VALVE TIME TO FAILURE AND THE IMPORTANCE OF STATISTICALLY SUPPORTED MAINTENANCE INTERVALS

Energy Technology Data Exchange (ETDEWEB)

Gross, R; Stephen Harris, S

2009-02-18

The Savannah River Site operates a Relief Valve Repair Shop certified by the National Board of Pressure Vessel Inspectors to NB-23, The National Board Inspection Code. Local maintenance forces perform inspection, testing, and repair of approximately 1200 spring-operated relief valves (SORV) each year as the valves are cycled in from the field. The Site now has over 7000 certified test records in the Computerized Maintenance Management System (CMMS); a summary of that data is presented in this paper. In previous papers, several statistical techniques were used to investigate failure on demand and failure rates including a quantal response method for predicting the failure probability as a function of time in service. The non-conservative failure mode for SORV is commonly termed 'stuck shut'; industry defined as the valve opening at greater than or equal to 1.5 times the cold set pressure. Actual time to failure is typically not known, only that failure occurred some time since the last proof test (censored data). This paper attempts to validate the assumptions underlying the statistical lifetime prediction results using Monte Carlo simulation. It employs an aging model for lift pressure as a function of set pressure, valve manufacturer, and a time-related aging effect. This paper attempts to answer two questions: (1) what is the predicted failure rate over the chosen maintenance/ inspection interval; and do we understand aging sufficient enough to estimate risk when basing proof test intervals on proof test results?

Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

International Nuclear Information System (INIS)

Li, D D; Jiang, J; Zhao, Z; Yi, W S; Lan, G

2013-01-01

We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system

Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

Science.gov (United States)

Li, D. D.; Jiang, J.; Zhao, Z.; Yi, W. S.; Lan, G.

2013-12-01

We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system.

Passive components of NPP safety-related systems

International Nuclear Information System (INIS)

Ionaytis Romuald, R.; Bubnova Tatyana, A.

2005-01-01

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

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.

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.

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

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

Preventive maintenance basis: Volume 16 -- Power operated relief valves, solenoid actuated. Final report

International Nuclear Information System (INIS)

Worledge, D.; Hinchcliffe, G.

1997-07-01

US nuclear plants are implementing preventive maintenance (PM) tasks with little documented basis beyond fundamental vendor information to support the tasks or their intervals. The Preventive Maintenance Basis project provides utilities with the technical basis for PM tasks and task intervals associated with 40 specific components such as valves, electric motors, pumps, and HVAC equipment. This report provides an overview of the PM Basis project and describes use of the PM Basis database. This volume 16 of the report provides a program of PM tasks suitable for application to power operated relief valves (PORV's) that are solenoid actuated. The PM tasks that are recommended provide a cost-effective way to intercept the causes and mechanisms that lead to degradation and failure. They can be used, in conjunction with material from other sources, to develop a complete PM program or to improve an existing program. Users of this information will be utility managers, supervisors, craft technicians, and training instructors responsible for developing, optimizing, or fine-tuning PM programs

Safety valve including a hydraulic brake and hydraulic brake that could be fitted into a valve

International Nuclear Information System (INIS)

Chabat-Courrede, Jean.

1981-01-01

Making of a safety valve that can be fitted to a containment vessel filled with a non compressible fluid, such as the water system of a nuclear power station. It includes a hydraulic brake located between the valve and the elastic means, close to the valve which completely suppresses the high frequency oscillations of the equipment [fr

Autonomous booster device of a safety valve

International Nuclear Information System (INIS)

Namand, H.

1983-01-01

The invention concerns an autonomous booster device of a protection safety valve of a pressure vessel. The valve comprises a hollow structure, a seat connected with a mobile flap forming one piece with a stem and a calibration spring bearing on the stem and on the valve structure to maintain the flap bearing on the seat. The stem of the flap is prolongated in a box forming one piece with the valve structure and receives an added push of a spring. The box acts as a pressure device of which the piston can exercise on the stem a push opposite to and larger than the spring one. The feeding device of the pressure box is finally described in detail [fr

Enhancement of pressurizer safety valve operability by seating design improvement

International Nuclear Information System (INIS)

Moisidis, N.T.; Ratiu, M.D.

1994-01-01

Operating conditions specific to Pressurizer Safety Valves (PSVs) have led to numerous problems and have caused industry and NRC concerns regarding the adequacy of spring loaded self-actuated safety valves for Reactor Coolant System (RCS) overpressure protection. Specific concerns are: setpoint drift, spurious actuations and leakage. Based on testing and valve construction analysis of a Crosby model 6M6 PSV, it was established that the primary contributor to the valve problems is a susceptibility to weak seating. To eliminate spring instability, a new spring washer was designed, which guides the spring and precludes its rotation from the reference installed position. Results of tests performed on a prototype PSV equipped with the modified upper spring washer has shown significant improvements in valve operability and a consistent setpoint reproducibility to less than ±1% of the PSV setpoint (testing of baseline, unmodified valve, resulted in a setpoint drift of ±2%). Enhanced valve operability will result in a significant decrease in operating and maintenance costs associated with valve maintenance and testing. In addition, the enhanced setpoint reproducibility will allow the development of a nitrogen to steam correlation for future in-house PSV testing which will result in further reductions in costs associated with valve testing

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.)

Safety Valve or Sinkhole? Vocational Schooling in South Africa

OpenAIRE

Pugatch, Todd

2012-01-01

As an alternative to traditional academic schooling, vocational schooling in South Africa may serve as a safety valve for students encountering difficulty in the transition from school to work. Yet if ineffective, vocational schooling could also be a sinkhole, offering little chance for success on the labor market. After defining the terms "safety valve" and "sinkhole" in a model of human capital investment with multiple schooling types, I test for evidence of these characteristics using a pa...

Increasing the Useful Life of Quench Reliefs with Inconel Bellows

Energy Technology Data Exchange (ETDEWEB)

Soyars, W. M. [Fermilab

1999-01-01

Reliable quench relief valves are an important part of superconducting magnet systems. Fermilab developed bellows-actuated cryogenic quench reliefs which have been in use since the early l 980's. The original design uses a stainless steel bellows. A high frequency, low amplitude vibration during relieving events has resulted in fatigue failures in the original design. To take advantage of the improved resistance to fatigue of Inconel, a nickel-chromium alloy, reliefs using Inconel 625 bellows were made. Design, development, and testing of the new version reliefs will be discussed. Tests show that relief valve lifetimes using Inconel bellows are more than five times greater than when using the original stainless steel bellows. Inconel bellows show great promise in increasing the lifetime of quench relief valves, and thus the reliability of accelerator cryogenic systems.

Prediction of two-phase choked-flow through safety valves

International Nuclear Information System (INIS)

Arnulfo, G; Bertani, C; De Salve, M

2014-01-01

Different models of two-phase choked flow through safety valves are applied in order to evaluate their capabilities of prediction in different thermal-hydraulic conditions. Experimental data available in the literature for two-phase fluid and subcooled liquid upstream the safety valve have been compared with the models predictions. Both flashing flows and non-flashing flows of liquid and incondensable gases have been considered. The present paper shows that for flashing flows good predictions are obtained by using the two-phase valve discharge coefficient defined by Lenzing and multiplying it by the critical flow rate in an ideal nozzle evaluated by either Omega Method or the Homogeneous Non-equilibrium Direct Integration. In case of non-flashing flows of water and air, Leung/Darby formulation of the two-phase valve discharge coefficient together with the Omega Method is more suitable to the prediction of flow rate.

Analysis on typical illegal events for nuclear safety class 1 valve

International Nuclear Information System (INIS)

Tian Dongqing; Gao Runsheng; Jiao Dianhui; Yang Lili; Chen Peng

2014-01-01

Illegal welding events of nuclear safety class l valve forging occurred to the manufacturer, while the valve was returned to be repaired. Illegal nondestructive test event of nuclear safety class valve occurred also to the manufacturer in the manufacturing process. The two events have resulted in quality incipient fault for the installed valves and the valves in the manufacturing process. It was reflected that operation of the factory quality assurance system isn't activated, and nuclear power engineering and operating company have insufficient supervision. The event-related parties should strengthen quality management and process control, get rid of the quality incipient fault, and experience feedback should be done well to guarantee quality of equipment in nuclear power plant. (authors)

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

Enhancement of pressurizer safety valve operability by seating design improvement

International Nuclear Information System (INIS)

Moisidis, N.T.; Ratiu, M.D.

1995-01-01

Operating conditions specific to pressurizer safety valves (PSVs) have led to numerous problems and have caused industry and NRC concerns regarding the adequacy of spring-loaded self-actuated safety valves for reactor coolant system (RCS) overpressure protection. Specific concerns are: setpoint drift, spurious actuations, and pressure protection. Specific concerns are: setpoint drift, spurious actuations, and leakage. Based on testing and valve construction analysis of a Crosby model 6M6 PSV (Moisidis and Ratiu, 1992), it was established that the primary contributor to the valve problems is a susceptibility to weak seating. To eliminate spring instability, a new spring washer was designed, which guides the spring and precludes its rotation from the reference installed position. Results of tests performed on a prototype PSV equipped with the modified upper spring washer has shown significant improvements in valve operability and a consistent setpoint reproducibility to less than ±1% of the PSV setpoint (testing of baseline, unmodified valve, resulted in a setpoint drift of ± 2%). Enhanced valve operability will result in a significant decrease in operating and maintenance costs associated with valve maintenance and testing. In addition, the enhanced setpoint reproducibility will allow the development of a nitrogen to steam correlation for future in-house PSV testing which will result in further reductions in costs associated with valve testing

Development of main steam safety valve set pressure evaluating system

International Nuclear Information System (INIS)

Oketani, Koichiro; Manabe, Yoshihisa.

1991-01-01

A main steam safety valve set pressure test is conducted for all valves during every refueling outage in Japan's PWRs. Almost all operations of the test are manually conducted by a skilled worker. In order to obtain further reliability and reduce the test time, an automatic test system using a personnel computer has been developed in accordance with system concept. Quality assurance was investigated to fix system specifications. The prototype of the system was manufactured to confirm the system reliability. The results revealed that this system had high accuracy measurement and no adverse influence on the safety valve. This system was concluded to be applicable for actual use. (author)

Analysis of high-pressure safety valves

NARCIS (Netherlands)

Beune, A.

2009-01-01

In presently used safety valve sizing standards the gas discharge capacity is based on a nozzle flow derived from ideal gas theory. At high pressures or low temperatures real gas effects can no longer be neglected, so the discharge coefficient corrected for flow losses cannot be assumed constant

To dimension safety valves. Probabilist study

International Nuclear Information System (INIS)

Noel, Robert; Couvreur, Denis

1982-01-01

The gauge of safety valves of a steam pressure apparatus is usually determined according to an operating situation envelope which it is admitted covers all that can happen in reality. For the safety of the dryer-superheaters of turbines in nuclear power stations, Electricite de France and Alsthom-Atlantique made a reliability study; its method is exposed and the results are discussed. Such a study is heavy going and complex, but in return it permits a better quantitative understanding of the various dimension and operating parameters of an installation which condition its safety. It is therefore a source of progress [fr

Water hammer caused by closure of turbine safety spherical valves

Science.gov (United States)

Karadžić, U.; Bergant, A.; Vukoslavčević, P.

2010-08-01

This paper investigates water hammer effects caused by closure of spherical valves against the discharge. During the first phase of modernisation of Perućica high-head hydropower plant (HPP), Montenegro, safety spherical valves (inlet turbine valves) have been refurbished on the first two Pelton turbine units. The valve closure is controlled by the valve actuator (hydraulic servomotor). Because the torque acting on the valve body is dependent on flow conditions the valve closing time may vary significantly for different flow velocities (passive valve). For the passive valve the torques acting on the valve body should be considered in the valve model. The valve closing time results from numerical simulation. On the contrary, for the active valve the valve closing time is assumed prior to simulation. The spherical valve boundary condition is incorporated into the method of characteristics (MOC) algorithm. The staggered (diamond) grid in applying the MOC is used in this paper. The passive valve boundary condition is described by the water hammer equations, the valve equation that relates discharge to pressure head drop and the dynamic equation of the valve body motion (torque equation). The active valve boundary condition is described by the first two equations, respectively. Standard quasi-steady friction model is used for estimating friction losses in plant's tunnel and penstocks. Numerical results using both the active and the passive spherical valve models are compared with results of measurements. It has been found that the influence of flow conditions on the spherical valve closing time is minor for the cases considered. Computed and measured results agree reasonably well.

Water hammer caused by closure of turbine safety spherical valves

International Nuclear Information System (INIS)

Karadzic, U; Vukoslavcevic, P; Bergant, A

2010-01-01

This paper investigates water hammer effects caused by closure of spherical valves against the discharge. During the first phase of modernisation of Perucica high-head hydropower plant (HPP), Montenegro, safety spherical valves (inlet turbine valves) have been refurbished on the first two Pelton turbine units. The valve closure is controlled by the valve actuator (hydraulic servomotor). Because the torque acting on the valve body is dependent on flow conditions the valve closing time may vary significantly for different flow velocities (passive valve). For the passive valve the torques acting on the valve body should be considered in the valve model. The valve closing time results from numerical simulation. On the contrary, for the active valve the valve closing time is assumed prior to simulation. The spherical valve boundary condition is incorporated into the method of characteristics (MOC) algorithm. The staggered (diamond) grid in applying the MOC is used in this paper. The passive valve boundary condition is described by the water hammer equations, the valve equation that relates discharge to pressure head drop and the dynamic equation of the valve body motion (torque equation). The active valve boundary condition is described by the first two equations, respectively. Standard quasi-steady friction model is used for estimating friction losses in plant's tunnel and penstocks. Numerical results using both the active and the passive spherical valve models are compared with results of measurements. It has been found that the influence of flow conditions on the spherical valve closing time is minor for the cases considered. Computed and measured results agree reasonably well.

Water hammer caused by closure of turbine safety spherical valves

Energy Technology Data Exchange (ETDEWEB)

Karadzic, U; Vukoslavcevic, P [Faculty of Mechanical Engineering, University of Montenegro Dzordza Vasingtona nn, Podgorica, 81000 (Montenegro); Bergant, A, E-mail: uros.karadzic@ac.m [LitostrojPower d.o.o., Litostrojska 50, Ljubljana, 1000 (Slovenia)

2010-08-15

This paper investigates water hammer effects caused by closure of spherical valves against the discharge. During the first phase of modernisation of Perucica high-head hydropower plant (HPP), Montenegro, safety spherical valves (inlet turbine valves) have been refurbished on the first two Pelton turbine units. The valve closure is controlled by the valve actuator (hydraulic servomotor). Because the torque acting on the valve body is dependent on flow conditions the valve closing time may vary significantly for different flow velocities (passive valve). For the passive valve the torques acting on the valve body should be considered in the valve model. The valve closing time results from numerical simulation. On the contrary, for the active valve the valve closing time is assumed prior to simulation. The spherical valve boundary condition is incorporated into the method of characteristics (MOC) algorithm. The staggered (diamond) grid in applying the MOC is used in this paper. The passive valve boundary condition is described by the water hammer equations, the valve equation that relates discharge to pressure head drop and the dynamic equation of the valve body motion (torque equation). The active valve boundary condition is described by the first two equations, respectively. Standard quasi-steady friction model is used for estimating friction losses in plant's tunnel and penstocks. Numerical results using both the active and the passive spherical valve models are compared with results of measurements. It has been found that the influence of flow conditions on the spherical valve closing time is minor for the cases considered. Computed and measured results agree reasonably well.

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.)

SEBIM pilot operated valves - CANDU and other applications

International Nuclear Information System (INIS)

Schaumburg, Gerald; Hera, Vlad

1999-01-01

The SEBIM Group, located at Chateauneuf-les-Martigues, near Marseilles, on the Mediterranean coast of France, is a market leader for the pilot operated safety relief valves in the nuclear industry. Its valves, which are subject to rigorous safety and reliability criteria, are intended to satisfy the most exigent requirements of not only the nuclear but also other demanding applications. The group manufacturing units are equipped with the state-of-the-art machinery, technical equipment and computer facilities. All personnel is highly specialized and trained. Among many applications the valves designed and manufactured by SEBIM for the biggest CANDU reactors in Canada were subjected to exceptionally difficult testing conditions, prior of being excepted by Ontario Hydro and passed all tests successfully. As a consequence Darlington N.G.S. was equipped with one of the most advanced SEBIM pressure protection piece of equipment, the Tandem Pilot Operated Pressure Relief Valve. Due to the demonstrated qualities of our product we were able to obtain the necessary registration of our valve original design with the appropriate Canadian authorities, both at the provincial level and at the federal level. One may find SEBIM protection and other type of equipment in civilian nuclear plants all over Europe as well as in military applications, like the French Navy. The SEBIM valves, covering a range between 15 mm and 160 mm diameter and capable of withstanding an inlet pressure from 2 MPa to 20 MPa and the temperature of the fluid up to 450 dec. C, are very strong competitors in the specialized field. Among these valves the tandems have special design and special qualities, for special applications. As mentioned above, two of these tandems are used, in parallel, on the Bleed Condenser vessels of the Darlington units to provide the ultimate protection of the Pressure and Inventory Control System and, through it, to the Main Heat Transfer System which is the primary cooling source

The safety valve and climate policy

International Nuclear Information System (INIS)

Jacoby, Henry D.; Ellerman, A.D.

2004-01-01

The 'safety valve' is a possible addition to a cap-and-trade system of emissions regulation whereby the authority offers to sell permits in unlimited amount at a pre-set price. In this way the cost of meeting the cap can be limited. It was proposed in the US as a way to control perceived high costs of the Kyoto Protocol, and possibly as a way to shift the focus of policy from the quantity targets of the Protocol to emissions price. In international discussions, the idea emerged as a proposal for a compliance penalty. The usefulness of the safety valve depends on the conditions under which it might be introduced. For a time it might tame an overly stringent emissions target. It also can help control the price volatility during the introduction of gradually tightening one, although permit banking can ultimately serve the same function. It is unlikely to serve as a long-term feature of a cap-and-trade system, however, because of the complexity of coordinating price and quantity instruments and because it will interfere with the development of systems of international emissions trade

Direct transcatheter aortic valve implantation with self-expandable bioprosthesis: Feasibility and safety

Energy Technology Data Exchange (ETDEWEB)

Fiorina, Claudia, E-mail: clafiorina@yahoo.it [Cardiac Catheterization Laboratory, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Maffeo, Diego; Curello, Salvatore [Cardiac Catheterization Laboratory, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Lipartiti, Felicia [Division of Cardiology, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Chizzola, Giuliano [Cardiac Catheterization Laboratory, Cardiothoracic Department, Spedali Civili, Brescia (Italy); D' Aloia, Antonio [Division of Cardiology, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Adamo, Marianna [Cardiac Catheterization Laboratory, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Mastropierro, Rosy [Division of Cardiothoracic Anestesiology, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Gavazzi, Emanuele [Department of Radiology, University of Brescia, Spedali Civili, Brescia (Italy); Ciccarese, Camilla; Chiari, Ermanna [Division of Cardiology, Cardiothoracic Department, Spedali Civili, Brescia (Italy); Ettori, Federica [Cardiac Catheterization Laboratory, Cardiothoracic Department, Spedali Civili, Brescia (Italy)

2014-06-15

Background: Balloon valvuloplasty has been considered a mandatory step of the transcatheter aortic valve implantation (TAVI), although it is not without risk. The aim of this work was to evaluate the feasibility and safety of TAVI performed without pre-dilation (direct TAVI) of the stenosed aortic valve. Material and Methods: Between June 2012 and June 2013, 55 consecutive TAVI performed without pre-dilation at our institution using the self-expandable CoreValve prosthesis (Medtronic, Minneapolis, MN) were analyzed and compared with 45 pre-dilated TAVI performed the previous year. Inclusion criteria were a symptomatic and severe aortic stenosis. Exclusion criteria were defined as presence of pure aortic regurgitation, degenerated surgical bioprosthesis or bicuspid aortic valve and prior procedure of balloon aortic valvuloplasty performed as a bridge to TAVI. Results: High-burden calcification in the device landing zone, assessed by CT scan, was found in most of the patients. The valve size implanted was similar in both groups. Device success was higher in direct TAVI (85% vs. 64%, p = 0.014), mostly driven by a significant lower incidence of paravalvular leak (PVL ≥2; 9% vs. 33%, p = 0.02). Safety combined end point at 30 days was similar in both groups. Conclusion: Compared to TAVI with pre-dilation, direct TAVI is feasible regardless of the presence of bulky calcified aortic valve and the valve size implanted. Device success was higher in direct TAVI, mostly driven by a lower incidence of paravalvular leak. Safety at 30 days was similar in two groups.

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.

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

IEEE Std 382-1980: IEEE standard for qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

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

International Nuclear Information System (INIS)

Koring, R.

2007-01-01

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

46 CFR 52.20-17 - Opening between boiler and safety valve (modifies PFT-44).

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Opening between boiler and safety valve (modifies PFT-44). 52.20-17 Section 52.20-17 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING POWER BOILERS Requirements for Firetube Boilers § 52.20-17 Opening between boiler and safety valve...

Efficacy and safety of the Lotus Valve System for treatment of patients with severe aortic valve stenosis and intermediate surgical risk

DEFF Research Database (Denmark)

De Backer, Ole; Götberg, Matthias; Ihlberg, Leo

2016-01-01

increasingly used to treat patients with an intermediate risk profile. METHODS AND RESULTS: The study was designed as an independent Nordic multicenter registry of intermediate risk patients treated with the Lotus Valve System (Boston Scientific, MA, USA; N=154). Valve Academic Research Consortium (VARC......)-defined device success was obtained in 97.4%. A Lotus Valve was successfully implanted in all patients. There was no valve migration, embolization, ectopic valve deployment, or TAV-in-TAV deployment. The VARC-defined combined safety rate at 30days was 92.2%, with a mortality rate of 1.9% and stroke rate of 3...

Safety significance of inadvertent operation of motor-operated valves in nuclear power plants

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.; Carbonaro, J.F.; Hall, R.E.

1994-01-01

Concerns about the consequences of valve mispositioning were brought to the forefront following an event at Davis Besse in 1985. The concern related to the ability to reposition open-quotes position-changeableclose quotes motor-operated valves (MOVs) from the control room in the event of their inadvertent operation and was documented in U.S. Nuclear Regulatory Commission (USNRC) Bulletin 85-03 and Generic Letter (GL) 89-10. The mispositioned MOVs may not be able to be returned to their required position due to high differential pressure or high flow conditions across the valves. The inability to reposition such valves may have significantly safety consequences, as in the Davis Besse event. However, full consideration of such mispositioning in safety analyses and in MOV test programs can be labor intensive and expensive. Industry raised concerns that consideration of position-changeable valves under GL 89-10 would not decrease the probability of core damage to an extent that would justify licensee costs. As a response, Brookhaven National Laboratory has conducted separate scoping studies for both boiling water reactors (BWRs) and pressurized water reactors (PWRs) using probabilistic risk assessment (PRA) techniques to determine if such valve mispositioning by itself is significant to safety. The approach used internal events PRA models to survey the order of magnitude of the risk-significance of valve mispositioning by considering the failure of selected position-changeable MOVs. The change in core damage frequency was determined for each valve considered, and the results were presented as a risk increase ratio for each of four assumed MOV failure rates. The risk increase ratios resulting from this failure rate sensitivity study can be used as a basis for a determination of the risk-significance of the MOV mispositioning issues for BWRs and PWRs

Comparison of EPRI safety valve test data with analytically determined hydraulic results

International Nuclear Information System (INIS)

Smith, L.C.; Howe, K.S.

1983-01-01

NUREG-0737 (November 1980) and all subsequent U.S. NRC generic follow-up letters require that all operating plant licensees and applicants verify the acceptability of plant specific pressurizer safety valve piping systems for valve operation transients by testing. To aid in this verification process, the Electric Power Research Institute (EPRI) conducted an extensive testing program at the Combustion Engineering Test Facility. Pertinent tests simulating dynamic opening of the safety valves for representative upstream environments were carried out. Different models and sizes of safety valves were tested at the simulated operating conditions. Transducers placed at key points in the system monitored a variety of thermal, hydraulic and structural parameters. From this data, a more complete description of the transient can be made. The EPRI test configuration was analytically modeled using a one-dimensional thermal hydraulic computer program that uses the method of characteristics approach to generate key fluid parameters as a function of space and time. The conservative equations are solved by applying both the implicit and explicit characteristic methods. Unbalanced or wave forces were determined for each straight run of pipe bounded on each side by a turn or elbow. Blowdown forces were included, where appropriate. Several parameters were varied to determine the effects on the pressure, hydraulic forces and timings of events. By comparing these quantities with the experimentally obtained data, an approximate picture of the flow dynamics is arrived at. Two cases in particular are presented. These are the hot and cold loop seal discharge tests made with the Crosby 6M6 spring-loaded safety valve. Included in the paper is a description of the hydraulic code, modeling techniques and assumptions, a comparison of the numerical results with experimental data and a qualitative description of the factors which govern pipe support loading. (orig.)

Study on the selection method of feed water heater safety valves in nuclear power plants

International Nuclear Information System (INIS)

Shi Jianzhong; Huang Chao; Hu Youqing

2014-01-01

The selection of the high pressure feedwater heater's safety valve usually follows the principle recommended by HEI standards in thermal power plant. However, the nuclear power plant's heaters generally need to accept a lots of drain from a moisture separator reheater (MSR). When the drain regulating valve was failure in fully open position, a large number of high pressure steam will directly goes into the heater. It make high-pressure heater have a risk of overpressure. Therefore, the safety valve selection of the heaters for nuclear power plants not only need to follow the HEI standards, but also need to check his capacity in certain special conditions. The paper established a calculation method to determine the static running point of the heaters based on characteristic equations of the feed water heater, drain regulating valve and steam extraction pipings, and energy balance principle. The method can be used to calculate the equilibrium pressure of various special running conditions, so further determine whether the capacity of the safety valve meets the requirements of safety and emissions. The method proposed in this paper not only can be used for nuclear power plants, can also be used for thermal power plants. (authors)

Steam relief valve control system for a nuclear reactor

International Nuclear Information System (INIS)

Torres, J.M.

1976-01-01

Described is a turbine follow system and method for Pressurized Water Reactors utilizing load bypass and/or atmospheric dump valves to provide a substitute load upon load rejection by bypassing excess steam to a condenser and/or to the atmosphere. The system generates a variable pressure setpoint as a function of load and applies an error signal to modulate the load bypass valves. The same signal which operates the bypass valves actuates a control rod automatic withdrawal prevent to insure against reactor overpower

Computational fluid dynamic simulation of pressurizer safety valve loop seal purge phenomena in nuclear power plants

International Nuclear Information System (INIS)

Park, Jong Woon

2012-01-01

In Korean 3 Loop plants a water loop seal pipe is installed containing condensed water upstream of a pressurizer safety valve to protect the valve disk from the hot steam environment. The loop seal water purge time is a key parameter in safety analyses for overpressure transients, because it delays valve opening. The loop seal purge time is uncertain to measure by test and thus 3-dimensional realistic computational fluid dynamics (CFD) model is developed in this paper to predict the seal water purge time before full opening of the valve which is driven by steam after water purge. The CFD model for a typical pressurizer safety valve with a loop seal pipe is developed using the computer code of ANSYS CFX 11. Steady-state simulations are performed for full discharge of steam at the valve full opening. Transient simulations are performed for the loop seal dynamics and to estimate the loop seal purge time. A sudden pressure drop higher than 2,000 psia at the tip of the upper nozzle ring is expected from the steady-state calculation. Through the transient simulation, almost loop seal water is discharged within 1.2 second through the narrow opening between the disk and the nozzle of the valve. It can be expected that the valve fully opens at least before 1.2 second because constant valve opening is assumed in this CFX simulation, which is conservative because the valve opens fully before the loop seal water is completely discharged. The predicted loop seal purge time is compared with previous correlation. (orig.)

Computational fluid dynamic simulation of pressurizer safety valve loop seal purge phenomena in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Nuclear and Energy Engineering Dept.

2012-11-15

In Korean 3 Loop plants a water loop seal pipe is installed containing condensed water upstream of a pressurizer safety valve to protect the valve disk from the hot steam environment. The loop seal water purge time is a key parameter in safety analyses for overpressure transients, because it delays valve opening. The loop seal purge time is uncertain to measure by test and thus 3-dimensional realistic computational fluid dynamics (CFD) model is developed in this paper to predict the seal water purge time before full opening of the valve which is driven by steam after water purge. The CFD model for a typical pressurizer safety valve with a loop seal pipe is developed using the computer code of ANSYS CFX 11. Steady-state simulations are performed for full discharge of steam at the valve full opening. Transient simulations are performed for the loop seal dynamics and to estimate the loop seal purge time. A sudden pressure drop higher than 2,000 psia at the tip of the upper nozzle ring is expected from the steady-state calculation. Through the transient simulation, almost loop seal water is discharged within 1.2 second through the narrow opening between the disk and the nozzle of the valve. It can be expected that the valve fully opens at least before 1.2 second because constant valve opening is assumed in this CFX simulation, which is conservative because the valve opens fully before the loop seal water is completely discharged. The predicted loop seal purge time is compared with previous correlation. (orig.)

Influence of safety vlave pressure on gelled electrolyte valve-regulated lead/acid batteries under deep cycling applications

International Nuclear Information System (INIS)

Oh, Sang Hyub; Kim, Myung Soo; Lee, Jin Bok; Lee, Heung Lark

2002-01-01

Cycle life tests have been carried out to evaluate the influence of safety valve pressure on vlave regulated lead/acid batteries under deep cycling applications. Batteries were cycled at 5 hour rates at 100 % DOD, and safety valve pressure was set to 1.08 and 2.00 bar, respectively. The batteries lost 248.3 g of water for each case after about 1,200 cycles, but the cyclic performances of the batteries were comparable. Most of the gas of the battery during discharging was hydrogen, and the oxygen concentration increased to 18 % after 3 hours of charging. The micro structure of the positive active materials was completely changed and the corrosion layer of the positive grid was less than 50 μm, regardless of the pressure of the safety valve after cycle life tests. The cause of discharge capacity decrease was found to water loss and the shedding of the positive active materials. The pressure of safety valve does not give little effect to the cyclic performance and the failure modes of the gelled electrolyte valve-regulated lead acid batteries

Phase separator safety valve blow-off.

CERN Multimedia

G. Perinic

2006-01-01

The fast discharge of the CMS solenoid leads to a pressure rise in the phase separator. On August 28th, a fast discharge was triggered at a current level of 19.1 kA. The pressure in the phase separator increased up to the set pressure of the safety valve and some helium was discharged. In consequence of this and prevoious similar observations the liquid helium level in the phase separator has been reduced from 60% to 50% and later to 45% in order to reduce the helium inventory in the magnet.

Ten-year rollover of San Onofre inservice testing program for pumps and valves to OM-6 and OM-10

International Nuclear Information System (INIS)

Croy, P.A.; Fischetti, S.; Chiang, D.; Schofield, P.; Barney, D.

1994-01-01

The Pump and Valve Inservice Testing (IST) Program Sat San Onofre, Units 2 and 3, was updated for the second 120-month interval from August 1993 to April 1994. The U.S. Nuclear Regulatory Commission (USNRC) approved the OM-6 and OM-10 Codes in mid-1992. The project for the rollover to these new Codes included several elements: (a) a review of the differences between IWV/IWP and OM-6/OM-10, (b) a comprehensive audit of the IST Program scope for valves, (c) creation of the program and supporting basis documents, the Relief Requests, and implementing procedures, (d) interdivisional coordination, (e) submittal to the USNRC, and (f) training. Subsections IWV and IWP have been used and essentially unchanged for over a decade. The new Code (Parts 1, 6, and 10 called OM-1, OM-6, and OM-10) includes several significant changes from the old Code. Our group identified these differences and drafted revised and reorganized Inservice Testing (IST) Program documents. We also considered USNRC Generic Letter 89-04 (GL 89-04), open-quotes Guidance on Developing Acceptable Inservice Testing Programsclose quotes, and NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, while revising the program. There were six pump relief requires and 13 valve relief requests in the program for the first 10-year interval. For the revised program we needed only one pump relief request (and no valve relief requests). Converting to the 1989 edition of the ASME Code did not require changes to the technical specifications. We revised our Updated Final Safety Analysis Report (UFSAR) to reflect the IST Program for the second 10-year interval. UFSAR changes were minor, consisting of updated references to the Code edition and 10 CFR 50.55a(f), open-quotes Inservice Testing Requirementsclose quotes

[Safety and efficacy of ketamine for pain relief].

Science.gov (United States)

Niesters, Marieke; Dahan, Albert; van Kleef, Maarten

2016-01-01

Intravenous ketamine treatment is frequently used for the management of chronic pain, especially in those patients who do not benefit from other therapies. In this commentary we discuss the efficacy of ketamine for relief of chronic pain and ketamine's safety profile. A review of the literature indicates that only a few studies show that intravenous ketamine has analgesic effects that persist beyond the infusion period, an effect that occurs in about two-thirds of patients. Ketamine has multiple safety issues, ranging from psychotomimetic and schizotypal symptoms, sympathetic stimulation, tachycardia and hypertension, and damage to the liver and the urogenital tract. Damage to the urogenital tract seems to be restricted to individuals who chronically abuse ketamine. We indicate the need for large randomized trials in which ketamine is compared with an 'active' placebo.

Conservative ground of qualification BRU-A VVER-1000 in modes of instability of diphasic environment

International Nuclear Information System (INIS)

Skalozubov, V.I.; Klyuchnikov, A.A.; Haj Farajallah Dabbach

2010-01-01

The article first presents grounds and conditions of origin of hydraulic shocks in the VVER system of safety relief valves, caused interchannel heat hydrodynamic instability of biphasic medium. It is supposed conservatively that origin of hydraulic shocks caused instability of biphasic stream determines the unavailability to close of safety relief valves. It is established that the modes of hydraulic shocks in safety relief valves of VVER 1000 (B-320) at the fully opened valves are not typical for the conditions of accidents with intercontour leakages.

Categorization of safety related motor operated valve safety significance for Ulchin Unit 3

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.

2002-03-01

We performed a categorization of safety related Motor Operated Valve (MOV) safety significance for Ulchin Unit 3. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure ( CCF) events in Ulchin Units 3 PSA. Therefore, in this study, we re-estimated the MGL(Multiple Greek Letter) parameter used for the evaluation of MOV CCF probabilities in Ulchin Units 3 Probabilistic Safety Assessment (PSA) and performed a classification of the MOV safety significance. The re-estimation results of the MGL parameter show that its value is decreased by 30% compared with the current value in Ulchin Unit 3 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter shows that the number of HSSCs(High Safety Significant Components) is decreased by 54.5% compared with those using the current value of it in Ulchin Units 3 PSA

Simulation of a hypothetical liquid relief valve failure (open) at Embalse nuclear power plant when a reactor shutdown is considered; Simulacion de la evolucion de la CNE (central nuclear Embalse) en el caso hipotetico de la apertura espuria de una valvula de alivio liquido con disparo del reactor

Energy Technology Data Exchange (ETDEWEB)

Bedrossian, G; Gersberg, S [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

1997-12-31

The study of the spurious opening of the liquid relief valves is of great interest in CANDU nuclear power plants because this could lead to a loss of coolant through the degasser-condenser relief valves, and implies an undesirable intermittent opening/closure of them. In fact, there is a specific procedure to follow at Embalse nuclear power plant whenever this abnormal situation occurs. This procedure contains a section where a reactor trip is considered. Really, automatic reactor trip is not accepted to occur. No trip parameters set points are through to be reached (neutronic or process). However, the procedure considers the situation where the reactor does trip. We analyzed the plant behavior when a reactor shutdown is triggered. Our objective was to assess if after this trip, the procedure can lead the plant to a safe situation, preventing high pressures in the degasser-condenser and with the inventory recovered in the storage tank. The case was analyzed with Firebird III, Mod. 1.0 code. Two situations were considered: trip at 40 sec. and trip at 180 sec. after the liquid relief valve failed opened (the latter when the degasser-condenser fills up). Procedure analysis and code simulations showed that following the steps recommended, provided the liquid relief valve can be closed manually, the inventory that enters the degasser-condenser from the heat transport primary system through the failed valve could be recovered in the storage tank, leading the plant to shutdown in safe conditions, and preventing the degasser-condenser relief valves setpoint from being reached. (author). 3 refs., 10 figs.

The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

2007-10-15

The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

Safety assessment of cast steel valve housing using NDE- and FM-methods

Energy Technology Data Exchange (ETDEWEB)

Blauel, J G; Hodulak, L [Fraunhofer-Institut fuer Werkstoffmechanik, Freiburg im Breisgau (Germany)

1988-12-31

This document presents some results concerning the safety assessment of cast steel valve housing. This risk assessment is performed through the Non Destructive Examination and Fracture Mechanics methods. (TEC).

Automated discovery of safety and efficacy concerns for joint & muscle pain relief treatments from online reviews.

Science.gov (United States)

Adams, David Z; Gruss, Richard; Abrahams, Alan S

2017-04-01

Product issues can cost companies millions in lawsuits and have devastating effects on a firm's sales, image and goodwill, especially in the era of social media. The ability for a system to detect the presence of safety and efficacy (S&E) concerns early on could not only protect consumers from injuries due to safety hazards, but could also mitigate financial damage to the manufacturer. Prior studies in the field of automated defect discovery have found industry-specific techniques appropriate to the automotive, consumer electronics, home appliance, and toy industries, but have not investigated pain relief medicines and medical devices. In this study, we focus specifically on automated discovery of S&E concerns in over-the-counter (OTC) joint and muscle pain relief remedies and devices. We select a dataset of over 32,000 records for three categories of Joint & Muscle Pain Relief treatments from Amazon's online product reviews, and train "smoke word" dictionaries which we use to score holdout reviews, for the presence of safety and efficacy issues. We also score using conventional sentiment analysis techniques. Compared to traditional sentiment analysis techniques, we found that smoke term dictionaries were better suited to detect product concerns from online consumer reviews, and significantly outperformed the sentiment analysis techniques in uncovering both efficacy and safety concerns, across all product subcategories. Our research can be applied to the healthcare and pharmaceutical industry in order to detect safety and efficacy concerns, reducing risks that consumers face using these products. These findings can be highly beneficial to improving quality assurance and management in joint and muscle pain relief. Copyright © 2017 Elsevier B.V. All rights reserved.

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

Energy Technology Data Exchange (ETDEWEB)

Sekhri, A.; Baumann, P. [KernkraftwerkLeibstadt AG, 5325 Leibstadt (Switzerland); Wicaksono, D. [Swiss Federal Inst. of Technology Zurich ETH, 8092 Zurich (Switzerland); Miro, R.; Barrachina, T.; Verdu, G. [Inst. for Industrial, Radiophysical and Environmental Safety ISIRYM, Universitat Politecnica de Valencia UPV, Cami de Vera s/n, 46021 Valencia (Spain)

2012-07-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

TRACE and TRAC-BF1 benchmark against Leibstadt plant data during the event inadvertent opening of relief valves

International Nuclear Information System (INIS)

Sekhri, A.; Baumann, P.; Wicaksono, D.; Miro, R.; Barrachina, T.; Verdu, G.

2012-01-01

In framework of introducing TRACE code to transient analyses system codes for Leibstadt Power Plant (KKL), a conversion process of existing TRAC-BF1 model to TRACE has been started within KKL. In the first step, TRACE thermal-hydraulic model for KKL has been developed based on existing TRAC-BF1 model. In order to assess the code models a simulation of plant transient event is required. In this matter simulations of inadvertent opening of 8 relief valves event have been performed. The event occurs at KKL during normal operation, and it started when 8 relief valves open resulting in depressurization of the Reactor Pressure Vessel (RPV). The reactor was shutdown safely by SCRAM at low level. The high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) have been started manually in order to compensate the level drop. The remaining water in the feedwater (FW) lines flashes due to saturation conditions originated from RPV depressurization and refills the reactor downcomer. The plant boundary conditions have been used in the simulations and the FW flow rate has been adjusted for better prediction. The simulations reproduce the plant data with good agreement. It can be concluded that the TRAC-BF1 existing model has been used successfully to develop the TRACE model and the results of the calculations have shown good agreement with plant recorded data. Beside the modeling assessment, the TRACE and TRAC-BF1 capabilities to reproduce plant physical behavior during the transient have shown satisfactory results. The first step of developing KKL model for TRACE has been successfully achieved and this model is further developed in order to simulate more complex plant behavior such as Turbine Trip. (authors)

Studies to demonstrate the adequacy of testing results of the qualification tests for the actuator of main steam safety relive valves (MSSRV) in an advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

Gou, P.F.; Patel, R.; Curran, G.; Henrie, D.; Solorzano, E.

2005-01-01

This paper presents several studies performed to demonstrate that the testing results from the qualification tests for the actuator of the Main Steam Safety Relief Valves (MSSRV; also called SRV in this paper) in GE's Advanced Boiling Water Reactor (ABWR) are in compliance with the qualification guidelines stipulated in the applicable IEEE standards. The safety-related function of the MSSRV is to relieve pressure in order to protect the reactor pressure vessel from over-pressurization condition during normal operation and design basis events. In order to perform this function, the SRV must actuate at a given set pressure while maintaining the pressure and structural integrity of the SRV. The valves are provided with an electro-pneumatic actuator assembly that opens the valve upon receipt of an automatic or manually initiated electric signal to allow depressurization of the reactor pressure vessel (RPV). To assure the SRV can perform its intended safety related functions properly, qualification tests are needed in addition to analysis, to demonstrate that the SRV can withstand the specified environmental, dynamic and seismic design basis conditions without impairing its safety related function throughout their installed life under the design conditions including postulated design basis events such as OBE loads and Faulted (SSE) events. The guidelines used for the test methods, procedures and acceptance criteria for the qualification tests are established in IEEE std 344-1987 and IEEE std 382-1985. In the qualification tests, the specimen consists of the actuator, control valve assembly, limit switches, and limit switch support structure. During the functional, dynamic and seismic tests, the test specimen was mounted on a SRV. Qualification of safety related equipment to meet the guidelines of the IEEE standards is typically a two-step process: 1) environmental aging and 2) design basis events qualification. The purpose of the first step is to put the equipment in an

On the Adequacy of API 521 Relief-Valve Sizing Method for Gas-Filled Pressure Vessels Exposed to Fire

Directory of Open Access Journals (Sweden)

Anders Andreasen

2018-03-01

Full Text Available In this paper, the adequacy of the legacy API 521 guidance on pressure relief valve (PRV sizing for gas-filled vessels subjected to external fire is investigated. Multiple studies show that in many cases, the installation of a PRV offers little or no protection—therefore provides an unfounded sense of security. Often the vessel wall will be weakened by high temperatures, before the PRV relieving pressure is reached. In this article, a multiparameter study has been performed taking into consideration various vessel sizes, design pressures (implicitly vessel wall thickness, vessel operating pressure, fire type (pool fire or jet fire by applying the methodology presented in the Scandpower guideline. A transient thermomechanical response analysis has been carried out to accurately determine vessel rupture times. It is demonstrated that only vessels with relatively thick walls, as a result of high design pressures, benefit from the presence of a PRV, while for most cases no appreciable increase in the vessel survival time beyond the onset of relief is observed. For most of the cases studied, vessel rupture will occur before the relieving pressure of the PRV is reached.

Analysis of operational methane emissions from pressure relief valves from biogas storages of biogas plants.

Science.gov (United States)

Reinelt, Torsten; Liebetrau, Jan; Nelles, Michael

2016-10-01

The study presents the development of a method for the long term monitoring of methane emissions from pressure relief valves (PRV(1)) of biogas storages, which has been verified during test series at two PRVs of two agricultural biogas plants located in Germany. The determined methane emission factors are 0.12gCH4kWhel(-1) (0.06% CH4-loss, within 106days, 161 triggering events, winter season) from biogas plant A and 6.80/7.44gCH4kWhel(-1) (3.60/3.88% CH4-loss, within 66days, 452 triggering events, summer season) from biogas plant B. Besides the operational state of the biogas plant (e.g. malfunction of the combined heat and power unit), the mode of operation of the biogas flare, which can be manually or automatically operated as well as the atmospheric conditions (e.g. drop of the atmospheric pressure) can also affect the biogas emission from PRVs. Copyright © 2016 Elsevier Ltd. All rights reserved.

VVER 1000 SBO calculations with pressuriser relief valve stuck open with ASTEC computer code

International Nuclear Information System (INIS)

Atanasova, B.P.; Stefanova, A.E.; Groudev, P.P.

2012-01-01

Highlights: ► We modelled the ASTEC input file for accident scenario (SBO) and focused analyses on the behaviour of core degradation. ► We assumed opening and stuck-open of pressurizer relief valve during performance of SBO scenario. ► ASTEC v1.3.2 has been used as a reference code for the comparison study with the new version of ASTEC code. - Abstract: The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC V2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) and Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), Germany. This investigation has been performed in the framework of the SARNET2 project (under the Euratom 7th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).

Design of emergency relief system to flare; Projeto de sistemas de alivio de emergencia para tocha

Energy Technology Data Exchange (ETDEWEB)

Britto, Joelson de Carvalho [PETROBRAS, Rio de Janeiro, RJ (Brazil)

2004-07-01

The relief system has for objective to protect the unit equipment and piping system for high pressures developed during eventual operational upset. As examples of operational upset could mention: human failure (operational mistake - example: inadvertent closure of a block valve), heat exchange tube rupture, utility failure (cooling water, electric power, steam, instrument air) and fire. The relieved products are piping to flare system in order to burn the waste gas. The burned or unburned relieved stream shall be dispersed in order to not to cause damages to the people and the environment. That system should operate automatically without the need of interference of the personnel of the operation. The system is basically constituted of Pressure Safety and Relief Valves (PSVs), piping net, gas-liquid separation vessel, separation vessel residual liquid pump (if necessary) and flare for burning waste gas without liquid. They are necessary also some utilities as fuel gas (to be used as purge gas by flare tip and as fuel gas by pilots in order to guarantee the continuous operation of the flare pilots), electric power, instrument/service air and compressed air or steam (if necessary) to improve the quality of the burns. (author)

Study of relief systems behavior in shipping terminals; Estudo do comportamento de sistemas de alivio em dutos de terminais maritimos

Energy Technology Data Exchange (ETDEWEB)

Pires, Luis F.G.; Ladeia, Renata C. da Cunha; Ortega Malca, Arturo J. [Pontificia Univ. Catolica do Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Mecanica. Nucleo de Simulacao Termohidraulica de Dutos (SIMDUT); Kossatz, Helmut [TRANSPETRO - PETROBRAS Transporte S.A., Rio de Janeiro, RJ (Brazil)

2005-07-01

In order to reduce operational costs, maritime terminals work with high flows transferring product between ships and in shore tanks. To protect the pipeline from any hydraulic transient, relief systems are commonly installed in strategic positions. During the project stage, computational simulation is widely used as auxiliary tool to determine the kind of valve and the place where the relief system must be installed. The present work points out the importance of the kind of valve chosen and its location along the pipeline, and shows that these factors affect directly the surge pressure produced by the valve closure. Besides, it compares a conventional spring-loaded relief valve behavior to a rupture disc one, making use of commercial program tools and employing a model based on the method of the characteristics. (author)

Aging and malfunction of valves in CANDU special safety systems. Phase 1

International Nuclear Information System (INIS)

1989-03-01

Aging and wear related valve malfunctions have been reported in American nuclear generating systems. This report documents the first attempt to study these phenomena on a global basis in Canadian nuclear power plants. A general methodology outlines an approach to this type of study which is amenable to use within existing information structures. Nuclear regulatory requirements which influence the testing of valves in Canadian nuclear power plants are reviewed. The reporting systems which emanate from these requirements are discussed and sources of valve failure data are reviewed. It is determined that modifications to existing failure reporting systems are required before practical means of collecting data necessary for the analysis of age related valve malfunctions can be developed. In spite of limitations in reported failure data, a partial data base is compiled for valve failures in Special Safety Systems of domestic nuclear plants. Data are reported for the period 1982 to 1986. The valve population and basic parameters of each valve such as type, operator, function, etc., and the reported failures against this population are compiled and reviewed for evidence of time dependent versus random failure trends. Results suggest that there is no clear age related failure trend. In fact, some systems and stations, experienced a reduction in failure rates with years of servicing, suggesting that some earlier generic valve problems may have been solved. Present inspection, test, and maintenance practices are reviewed and their effectiveness for purposes of predicting or preventing incipient failures is assessed to be of moderate value. Modern failure prevention methods are highlighted and their applicability discussed

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.

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.

Numerical simulation on flow field of nuclear safety grade 2 single-seat pneumatic diaphragm control valve

International Nuclear Information System (INIS)

Zhong Yun; Zhang Jige; Wang Dezhong; Shi Jianzhong

2010-01-01

The Computational Fluid Dynamics (CFD) method is employed to simulate numerically the steady flow and transient flow under variable openings of the nuclear safety grade 2 single-seat pneumatic diaphragm control valve, which is a sleeve valve. The steady simulations under rated condition tells that there is a large amount of vortex in the valve seat necking and around the valve cone, which leads to a much greater flow impact on the head of the valve cone and uneven pressure distribution on spool face. More consideration should be taken on the characteristics of the valve cone accordingly, when designing a valve of this kind. Then the transient flow under 100% and 40% openings is simulated numerically on the basis of steady simulations. The pulsation of the pressure magnitude at the points with large vorticity, in the valve seat necking and around the valve cone, is monitored. The main pulsation frequencies differ from the low natural frequencies of the model, which means that it is safe from leading to structural resonance. (authors)

A study on the development of the computerized safety evaluation system of the motor operated valve

International Nuclear Information System (INIS)

Kim, J. C.; Park, S. G.; Lee, D. H.; Ahn, N. S.; Bae, H. J.; Hong, J. S.

2001-01-01

The MOVIDIK (Motor-Operated Valves Integrated Database and Information of KEPCO) system was developed to assist the design basis safety evaluation and to manage the overall data made by evaluation on the safety-related Motor-operated Valves(MOV) in the nuclear power plant. The huge amount of safety evaluation data of the MOV is being piled up as the safety evaluation work goes on. Much time and manpower was needed to do safety evaluation works without computerized system and it was not easy to obtain the statistic information from the evaluation data. The MOVIDIK will improve the efficiency of safety evaluation works and standardize the analysis process. But the some process which needs specific evaluation codes and engineering calculation by the specialists was not computerized. The MOVIDIK was developed by JAVA/JSP language known by the flexibility of language and the easiness of transplantation between operating systems. The Oracle 8i which is the world's most popular database was used for MOVIDIK database

SAFETY ALERT - Failure of brass non-return valves in gas point installations

CERN Multimedia

HSE Unit

2016-01-01

There have been three recent failures in brass non-return valves in separate high pressure gas point installations across CERN. Whilst each was in a different gas service, the visual nature of the failure has been similar.   In all three cases, these components were connected to stainless steel flexible connections and stainless steel pipework. From the metallurgical investigation of the failed component, it appears that the failure is linked to uncontrolled tightening, leading to a localised weakening resulting in premature failure when subjected to pressure. Lead levels in the examined components appear to be a contributing factor to the reduction in ductility but are not identified as the root cause. It has also not been possible to attribute failure to a particular batch of material. The Occupational Health & Safety and Environmental Protection Unit prescribes the following actions to be taken, aligned with the CERN Safety Rules: Verification of all brass non-return valves (prioritising...

PWR pressurizer discharge piping system on-site testing

International Nuclear Information System (INIS)

Anglaret, G.; Lasne, M.

1983-08-01

Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

AREVA's innovative solutions for valve diagnostics and in-situ valve repair

International Nuclear Information System (INIS)

Damies, H.; Breitenberger, U.; Munoz, L.; Kostroun, F.

2012-01-01

Optimized maintenance strategies are a key aspect for safe and undisturbed plant operation. Innovative valve service solutions can support that in an efficient way. The ADAM®/SIPLUG® valve monitoring system allows full online monitoring of valves and actuators with automatic evaluation and assessment. Especially for safety-related and operation-related valves this provides valuable information on components condition to ensure proper function and contribute to optimization of maintenance strategies as well as effective maintenance performance. More than 25 years of experience in various plants worldwide show that application of ADAM®/SIPLUG® valve diagnostics solution leads to increased plant safety and availability. With the innovative AVARIS technology an in-situ valve repair is possible. It has the unique ability to conduct several steps in-situ, to maintain the sealing seat of gate or check valves. By applying AVARIS, the valve is restored in its original state, the system remains unchanged. Thus, all original documents remain valid and applicable. In comparison to previous procedures like cutting valves out of the pipeline and repairing hard facings or damaged seal seats in a separate workshop or alternatively replacement by a new valve body the new AVARIS technology avoids costs, risk and effort. (author)

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.

Dynamic simulation of relief line during loss of insulation vacuum of the ITER cryoline

Science.gov (United States)

Badgujar, S.; Kosek, J.; Grillot, D.; Forgeas, A.; Sarkar, B.; Shah, N.; Choukekar, K.; Chang, H.-S.

2017-12-01

The ITER cryoline (CL) system consists of 37 types of vacuum jacketed transfer lines which forms a complex structured network with a total length of about 5 km, spread inside the Tokamak building, on a dedicated plant bridge and in the Cryoplant building/area. One of them, the low pressure relief line (RL) recovers helium discharged from process safety relief valves of the different cryogenic users and is sent it back to the Cryoplant via heater and recovery system. The process pipe diameters of the RL vary from DN 50 to DN 200 and the length is more than 1500 m. Loss of insulation vacuum (LIV) of a CL is one of the worst scenarios apart from LIV in Auxiliary Cold Boxes (ACBs). The Torus and Cryostat CL is chosen to simulate the virtual LIV and to study the anticipated behavior of the RL. Both helium LIV (LIV due to leak in helium pipe) and air LIV (LIV due to air ingress in outer vacuum jacket of the cryoline) with and without fire) have been simulated during this study. After the brief description of the CL system, the paper will describe the EcosimPro® model prepared for the dynamic study. The paper will also describe the results like minimum temperature of RL, mass flow and maximum pressure in the RL which are essentially used to choose the type and location of safety relief devices to protect the CL process pipes.

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...

The numerical evaluation of the minimal outlet area of the safety valve in the pipelines of cryogenic installations

CERN Multimedia

CERN. Geneva

2016-01-01

The flow of cold helium in pipes is a fundamental issue of any cryogenic installation. Pipelines for helium transportation can reach lengths of hundreds of meters. The proper selection of size for individual pipelines and safety valves is a crucial part in the consideration of costs for the entire installation and its safe operation. The size of the safety valve must be properly designed in order to avoid a dangerous pressure buildup during normal operation, as well as in the case of emergency. The most commonly occurring dangerous situation is an undesired heat flux in the helium as a result of a broken insulation. In this case, the heat flux can be very intense and the buildup of the pressure in the pipe can be very rapid. In the present work, numerical calculations were used to evaluate the buildup of pressure and temperature in the pipe, in the case of a sudden and intense heat flux. The main goal of the applied numerical procedure was to evaluate the proper sizes of the safety valves in order to avoid a...

Simulant Development for Hanford Tank Farms Double Valve Isolation (DVI) Valves Testing

Energy Technology Data Exchange (ETDEWEB)

Wells, Beric E.

2012-12-21

Leakage testing of a representative sample of the safety-significant isolation valves for Double Valve Isolation (DVI) in an environment that simulates the abrasive characteristics of the Hanford Tank Farms Waste Transfer System during waste feed delivery to the Waste Treatment and Immobilization Plant (WTP) is to be conducted. The testing will consist of periodic leak performed on the DVI valves after prescribed numbers of valve cycles (open and close) in a simulated environment representative of the abrasive properties of the waste and the Waste Transfer System. The valve operations include exposure to cycling conditions that include gravity drain and flush operation following slurry transfer. The simulant test will establish the performance characteristics and verify compliance with the Documented Safety Analysis. Proper simulant development is essential to ensure that the critical process streams characteristics are represented, National Research Council report “Advice on the Department of Energy's Cleanup Technology Roadmap: Gaps and Bridges”

77 FR 5472 - Pipeline Safety: Expanding the Use of Excess Flow Valves in Gas Distribution Systems to...

Science.gov (United States)

2012-02-03

..., Regulatory Certainty, and Job Creation Act of 2011 (PL112-90), have imposed additional demands on their... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration 49 CFR Part 192 [Docket ID PHMSA-2011-0009] RIN 2137-AE71 Pipeline Safety: Expanding the Use of Excess Flow Valves...

Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Chang Hyun; Kim, Young Ae; Oh, Seung Jong; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

2007-10-15

The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

Directly acting spring loaded safety valves as shock reducing measure

International Nuclear Information System (INIS)

Ismaier, A.; Schluecker, E.

2010-01-01

Hydraulic shocks as induced by fast closure of armatures or by sudden pump failures are massive impacts in piping systems and require extensive measures to absorb the generated load. Basically the avoidance of water hammers are preferable but in case of emergency shutdowns unavoidable hydraulic shocks have to be reduced by appropriate measures. The authors describe experiments with spring loaded safety valves as shock reducing measures. It was shown that the vale dimensions is essential for the efficacy. A realistic modeling is possible using the one-dimensional fluid mechanics code ROLAST.

Pressure locking and thermal binding of gate valves

Energy Technology Data Exchange (ETDEWEB)

Kelly, E.M.

1996-12-01

Pressure locking and thermal binding represent potential common mode failure mechanisms that can cause safety-related power-operated gate valves to fail in the closed position, thus rendering redundant safety-related systems incapable of performing their safety functions. Supplement 6 to Generic Letter 89-10, {open_quotes}Safety-Related Motor-Operated Gate Valve Testing and Surveillance,{close_quotes} provided an acceptable approach to addressing pressure locking and thermal binding of gate valves. More recently, the NRC has issued Generic Letter 95-07, {open_quotes}Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,{close_quotes} to request that licensees take certain actions to ensure that safety-related power-operated gate valves that are susceptible to pressure locking or thermal binding are capable of performing their safety functions within the current licensing bases. Over the past two years, several plants in Region I determined that valves in certain systems were potentially susceptible to pressure locking and thermal binding, and have taken various corrective actions. The NRC Region I Systems Engineering Branch has been actively involved in the inspection of licensee actions in response to the pressure locking and thermal binding issue. Region I continues to maintain an active involvement in this area, including participation with the Office of Nuclear Reactor Regulation in reviewing licensee responses to Generic Letter 95-07.

Nuclear safety in Slovak Republic. Status of safety improvements

International Nuclear Information System (INIS)

Toth, A.

1999-01-01

Status of the safety improvements at Bohunice V-1 units concerning WWER-440/V-230 design upgrading were as follows: supplementing of steam generator super-emergency feed water system; higher capacity of emergency core cooling system; supplementing of automatic links between primary and secondary circuit systems; higher level of secondary system automation. The goal of the modernization program for Bohunice V-1 units WWER-440/V-230 was to increase nuclear safety to the level of the proposals and IAEA recommendations and to reach probability goals of the reactor concerning active zone damage, leak of radioactive materials, failures of safety systems and damage shields. Upgrading program for Mochovce NPP - WWER-440/V-213 is concerned with improving the integrity of the reactor pressure vessel, steam generators 'leak before break' methods applied for the NPP, instrumentation and control of safety systems, diagnostic systems, replacement of in-core monitoring system, emergency analyses, pressurizers safety relief valves, hydrogen removal system, seismic evaluations, non-destructive testing, fire protection. Implementation of quality assurance has a special role in improvement of operational safety activities as well as safety management and safety culture, radiation protection, decommissioning and waste management and training. The Year 2000 problem is mentioned as well

Safety analysis of Oi nuclear power plant

International Nuclear Information System (INIS)

1979-01-01

The transient phenomena in Oi nuclear power plant were analyzed, especially on the water level fluctuation and the capability of natural circulation in the primary loop, under the assumptions that the feed water for steam generators is totally lost, and the relief valve on the pressurizer, which is actuated due to the pressure rise in the primary system, is stuck and kept open. These assumptions are related to the TMI accident. The analysing conditions are 1) the main feed water flow is totally lost suddenly during the rated power operation of the reactor, 2) two motor-driven auxiliary feed water pumps are started manually fifteen minutes after the accident initiation, 3) one relief valve on the pressurizer is opened fifteen seconds after the accident initiation and kept open, 4) the reactor is scrammed thirty three seconds after the accident initiation, 5) the turbine is tripped 33.5 seconds after the accident initiation, etc. Two cases were analysed, namely 3,800 seconds and 1,200 seconds after the accident initiation. The analytical code RELEP4/Mod5/U2/J1 was utilized for this analysis. The level fluctuation in the pressurizer after the accident initiation, the flow rate fluctuation through the pressurizer relief valve, especially that of steam, liquid single phase and two phase flows, the water level in the upper plenum in the pressure vessel, the change of flow rate at core inlet, the average pressure in the core, and the temperature fluctuation of coolant in the core, the variation of void fraction in the core, and the change of surface temperature of fuel rods are presented as the analysis results, and they are evaluated. It is recognized that the plant safety is kept under the assumed accident conditions in the Oi nuclear power plant. (Nakai, Y.)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Maximizing prosthetic valve size with the Top Hat supra-annular aortic valve

DEFF Research Database (Denmark)

Aagaard, Jan; Geha, Alexander S.

2007-01-01

BACKGROUND AND AIM OF THE STUDY: The CarboMedics Top Hat supra-annular aortic valve allows a one-size (and often two-size) increase over the standard intra-annular valve. This advantage should minimize the risk of patient-prosthesis mismatch, where the effective prosthetic valve orifice area....... This study evaluates the authors' clinical experience with Top Hat supra-annular aortic valve size selection, and the technical aspects of implantation. METHODS: Between January 1999 and October 2005, a total of 251 consecutive patients underwent 252 aortic valve replacements with Top Hat supra...... required unplanned coronary bypass, and 30-day mortality was 2.0% (5/251), indicating a good safety profile for the valves implanted in this series. CONCLUSION: The general distribution of implant sizes in the US indicates that cardiac surgeons may be under-sizing the Top Hat supra-annular aortic valve...

Pilot program to identify valve failures which impact the safety and operation of light water nuclear power plants

International Nuclear Information System (INIS)

Tsacoyeanes, J.C.; Raju, P.P.

1980-04-01

The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents

Mitral valve surgery using right anterolateral thoracotomy: is the aortic cannulation a safety procedure?

Science.gov (United States)

Guedes, Marco Antonio Vieira; Pomerantzeff, Pablo Maria Alberto; Brandão, Carlos Manuel de Almeida; Vieira, Marcelo Luiz Campos; Grinberg, Max; Stolf, Noedir Antonio Groppo

2010-01-01

The right anterolateral thoracotomy is an alternative technique for surgical approach of mitral valve. In these cases, femoral-femoral bypass still has been used, rising occurrence of complications related to femoral cannulation. Describe the technique and results of mitral valve treatment by right anterolateral thoracotomy using aortic cannulation for cardiac pulmonary bypass (CPB). From 1983 e 2008, 100 consecutive female patients, with average age 35 ±13 years, 96 (96%) underwent mitral valve surgical treatment in the Heart Institute of São Paulo. A right anterolateral thoracotomy approach associated with aortic cannulation was used for CPB. Eighty (80%) patients had rheumatic disease and 84 (84%) patients presented functional class III or IV. Were performed 45 (45%) comissurotomies, 38 (38%) valve repairs, 7(7%) mitral valve replacements, seven (7%) recomissurotomies and three (3%) prosthesis replacement. Sparing surgery was performed in 90 (90%) patients. The average CPB and clamp time were 57 ± 27 min e 39 ± 19 min, respectively. There were no in-hospital death, reoperation due to bleeding and convertion to sternotomy. Introperative complications were related to heart harvest (5%), especially in reoperations (3%). The most important complications in postoperative period were related to pulmonary system (11%), followed by atrial fibrilation (10%) but without major systemic repercussions. The mean inhospital length of stay was 8 ± 3 days. Follow-up was 6.038 patients/month. Actuarial survival was 98.0 ± 1.9% and freedom from reoperation was 81.4 ± 7.8% in 180 months. The right anterolateral thoracotomy associated with aortic cannulation in mitral valve surgery is a simple technique, reproducible and safety.

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

Experimental investigation of processes in primary circuit relief system

International Nuclear Information System (INIS)

Tomas, Z.; Simo, T.; Konecny, A.

1989-01-01

The protective condenser (direct contact condenser) is one of the basic components of the primary circuit relief system of WWER power plants. The steam flowing from the surge tank through relief valves into the subcooled water condensates in the protective condenser vessel. Two simple physical models were designed and constructed for investigation of bubbling through (contact condensation). An experimental program was performed with the aim of determining the distribution of temperatures in the axis of the steam jet and its vicinity, determining the velocity field of water into vicinity of steam jets, observing the geometrical shape of jets and their interaction and determining important values for mathematical model. (orig.)

Regulatory control of nuclear facility valves and their actuators

International Nuclear Information System (INIS)

1993-01-01

The methods and procedures by which the Finnish Centre for Radiation and Nuclear Safety (STUK) regulates valves and their actuators in nuclear power plants and in other nuclear facilities are specified in the guide. The scope of regulation depends on the Safety Class of the valve and the actuator in question. The Safety Classification principles for the systems, structures and components of the nuclear power plants are described in the guide YVL 2.1 and the regulatory control of the nuclear facility safety valves is described in the guide YVL 5.4

Application of risk-based methods to inservice testing of check valves

Energy Technology Data Exchange (ETDEWEB)

Closky, N.B.; Balkey, K.R.; McAllister, W.J. [and others

1996-12-01

Research efforts have been underway in the American Society of Mechanical Engineers (ASME) and industry to define appropriate methods for the application of risk-based technology in the development of inservice testing (IST) programs for pumps and valves in nuclear steam supply systems. This paper discusses a pilot application of these methods to the inservice testing of check valves in the emergency core cooling system of Georgia Power`s Vogtle nuclear power station. The results of the probabilistic safety assessment (PSA) are used to divide the check valves into risk-significant and less-risk-significant groups. This information is reviewed by a plant expert panel along with the consideration of appropriate deterministic insights to finally categorize the check valves into more safety-significant and less safety-significant component groups. All of the more safety-significant check valves are further evaluated in detail using a failure modes and causes analysis (FMCA) to assist in defining effective IST strategies. A template has been designed to evaluate how effective current and emerging tests for check valves are in detecting failures or in finding significant conditions that are precursors to failure for the likely failure causes. This information is then used to design and evaluate appropriate IST strategies that consider both the test method and frequency. A few of the less safety-significant check valves are also evaluated using this process since differences exist in check valve design, function, and operating conditions. Appropriate test strategies are selected for each check valve that has been evaluated based on safety and cost considerations. Test strategies are inferred from this information for the other check valves based on similar check valve conditions. Sensitivity studies are performed using the PSA model to arrive at an overall IST program that maintains or enhances safety at the lowest achievable cost.

Thermal overload protection for electric motors on safety-related motor-operated valves: Generic Issue II.E.6.1

International Nuclear Information System (INIS)

Rothberg, O.

1988-06-01

NRC regulatory positions, as stated in Regulatory Guide 1.106, Revision 1, have been identified by the Office for Analysis and Evaluation of Operational Data (AEOD) as potential contributors to valve motor burnout. AEOD is particularly concerned about the allowed policy of bypassing thermal overload devices during normal or accident conditions. Regulatory Guide 1.106 favors compromising the function of thermal overload devices in favor of completing the safety-related action of valves. The purpose of this study was to determine if the guidance contained in Regulatory Guide 1.106 is appropriate and, if not, to recommend the necessary changes. This report describes thermal overload devices commonly used to protect safety-related valve operator motors. The regulatory guidelines stated in Regulatory Guide 1.106 along with the limitations of thermal overload protection are discussed. Supplements and alternatives to thermal overload protection are also described. Findings and conclusions of several AEOD reports are discussed. Information obtained from the standard review plan, standard technical specifications, technical specifications from representative plants, and several papers are cited

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.)

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

Directory of Open Access Journals (Sweden)

In-Bok Lee

2016-12-01

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

IEEE Std 382-1985: IEEE standard for qualification of actuators for power operated valve assemblies with safety-related functions for nuclear power plants

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of power-driven valve actuators with safety-related functions Part I describes the qualification process. Part II describes the standard qualification cases and their environmental parameters for the usual locations of safety-related equipment in a nuclear generating station. Part III describes the qualification tests outlined in 6.3.3

10 Years of operating experience of the valves in the safety systems on Caorso plant

International Nuclear Information System (INIS)

Curcuruto, S.; Pasquini, M.

1990-01-01

The Operating Experience (O.E.) of the valves in the safety related systems on Caorso plant has been analysed. The valves have been grouped according to system, type and manufacturer. All the data on the failures have been respectively drawn out by the O.E. data bank and, in some cases, they have been integrated by informations collected directly on the plant. The events and the relevant causes have been analysed, particularly taking into account the repetitive events. Most of the failures were discovered during the surveillance tests, giving a positive indication of the effectiveness of the periodic test program. It was also that concluded hardware problems caused more failures than human errors both during operation and maintenance. Abnormal distributions of failures on the valves and on their components have been found out. Weak components both mechanical and electrical and pertinent corrective measures have been identified, aimed to eliminate the recurring failure modes

International Environmental Agreements: Emissions Trade, Safety Valves and Escape Clauses

International Nuclear Information System (INIS)

Karp, Larry; Zhao, Jinhua

2010-01-01

We explain how the structure of multi-national or multi-regional environmental agreements affect their chance of success. Trade in emissions permits has ambiguous and in some cases surprising effects on both the equilibrium level of abatement, and on the ability to persuade nations or regions to participate in environmental agreements. An escape clause policy and a safety valve policy have essentially the same properties when membership in environmental agreement is pre-determined, but they create markedly different effects on the incentives to join such an agreement. The two policies lead to a qualitative difference in the leverage that a potential member of the agreement exercises on other members

Trial evaluations in comparison with the 1983 safety goals

International Nuclear Information System (INIS)

Riggs, R.; Sege, G.

1985-06-01

This report provides retrospective comparisons of selected generic regulatory actions to the 1983 NRC safety goals, which had been issued for evaluation during a two-year period. The issues covered are those analyzed by the Office of Nuclear Reactor Regulation (NRR) (assisted in some cases by the Battelle Pacific Northwest Laboratory). The issues include auxiliary feedwater reliability, pressurized thermal shock, power-operated relief valve isolation, asymmetric blowdown loads on PWR primary systems, pool dynamic loads for BWR containments, and steam generator tube rupture. Calculated core-melt frequencies, mortality risks, and cost-benefit ratios are compared with the corresponding safety-goal quantitative design objectives. Considerations that should influence interpretation of the comparisons are discussed. Comments are included on whether and how the safety goals may have helped in the regulatory decision process and on problems encountered

Approach for a modeling extension for relief valves in one-dimensional calculation codes with respect to the evaluation of water hammer effects in piping systems; Ansatz zur Erweiterung der Modellierung von Rueckschlagklappen in 1-D Rechencodes hinsichtlich der Bewertung von Druckstoessen in Rohrleitungssystemen

Energy Technology Data Exchange (ETDEWEB)

Frings, Malte; Malcher, Daniel [Westinghouse Electric Germany GmbH, Mannheim (Germany)

2012-11-01

Relief valves are used in industrial plants for instance as safeguarding of the system pressure in case of pump failures. Pump failures and automatic changeovers to redundant aggregates cause a flow reversal which induces the stop valve closure. This process can cause water hammer effects in the piping system. The backflow velocity defines the maximum load in the piping system. The presented approach taking into account this effect of medium displacement in the RELAP calculations yields significant differences to the former results. Validation using experimental data is required.

Stretchable Spin Valve with Stable Magnetic Field Sensitivity by Ribbon-Patterned Periodic Wrinkles.

Science.gov (United States)

Li, Huihui; Zhan, Qingfeng; Liu, Yiwei; Liu, Luping; Yang, Huali; Zuo, Zhenghu; Shang, Tian; Wang, Baomin; Li, Run-Wei

2016-04-26

A strain-relief structure by combining the strain-engineered periodic wrinkles and the parallel ribbons was employed to fabricate flexible dual spin valves onto PDMS substrates in a direct sputtering method. The strain-relief structure can accommodate the biaxial strain accompanying with stretching operation (the uniaxial applied tensile strain and the induced transverse compressive strain due to the Poisson effect), thus significantly reducing the influence of the residual strain on the giant magnetoresistance (GMR) performance. The fabricated GMR dual spin-valve sensor exhibits the nearly unchanged MR ratio of 9.9%, magnetic field sensitivity up to 0.69%/Oe, and zero-field resistance in a wide range of stretching strain, making it promising for applications on a conformal shape or a movement part.

46 CFR 56.01-1 - Scope (replaces 100.1).

Science.gov (United States)

2010-10-01

... ships' and barges' piping systems and appurtenances. (b) The respective piping systems installed on ships and barges shall have the necessary pumps, valves, regulation valves, safety valves, relief valves... operation of the vessel. (c) Piping for industrial systems on mobile offshore drilling units need not fully...

Percutaneous implantation of the CoreValve aortic valve prosthesis in patients at high risk or rejected for surgical valve replacement: Clinical evaluation and feasibility of the procedure in the first 30 patients in the AMC-UvA

NARCIS (Netherlands)

Baan, J.; Yong, Z. Y.; Koch, K. T.; Henriques, J. P. S.; Bouma, B. J.; de Hert, S. G.; van der Meulen, J.; Tijssen, J. G. P.; Piek, J. J.; de Mol, B. A. J. M.

2010-01-01

Objective. To report the feasibility, safety and efficacy of percutaneous aortic valve implantation (PAVI) with the CoreValve self-expanding aortic valve bioprosthesis in elderly patients with aortic valve stenosis who are rejected for surgery or have a high surgical risk.Methods. PAVI using the

46 CFR 154.801 - Pressure relief systems.

Science.gov (United States)

2010-10-01

... 46 Shipping 5 2010-10-01 2010-10-01 false Pressure relief systems. 154.801 Section 154.801 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY... Vent Systems § 154.801 Pressure relief systems. (a) Each cargo tank that has a volume of 20m3 (706 ft.3...

76 FR 72666 - Pipeline Safety: Expanding the Use of Excess Flow Valves in Gas Distribution Systems to...

Science.gov (United States)

2011-11-25

... technical feasibility and cost of the installation of such valves; (D) The public safety benefits of the... public comment regarding the technical challenges, and the potential costs and the potential benefits of... a cost-benefit perspective. DATES: Persons interested in submitting written comments on this ANPRM...

Nuclear valves and the licencing authority

International Nuclear Information System (INIS)

Stevens-Guille, P.D.

1977-02-01

Valve problems relevant to licencing authorities are discussed. It is suggested that the probability of valve failure should be given more emphasis in safety analysis. Problems of stress analysis, codes, seismic effects, malfunction and leakage are discussed and programmes aimed at solving future problems are outlined

A sigh of relief or a sigh to relieve: The psychological and physiological relief effect of deep breaths.

Science.gov (United States)

Vlemincx, Elke; Van Diest, Ilse; Van den Bergh, Omer

2016-10-15

Both animal and human research have revealed important associations between sighs and relief. Previously we argued to conceive of sighs as resetters which temporarily induce relief. The present study aimed to investigate the psychological and physiological relief effect of sighs by instructed deep breaths and spontaneous sighs compared to a control breathing maneuver. Participants completed three blocks of 40 trials during which uncertainty cues were followed by either safety cues followed by a positive picture, or danger cues followed by a negative picture. One block was presented without breathing instructions, two subsequent blocks with breathing instructions. During the presentation of the safety and danger cues, an instruction was given to either 'take a deep breath' or 'postpone the next inhalation for 2 s (breath hold). Continuously, participants rated relief and Frontalis electromyography was recorded. Trait anxiety sensitivity was assessed by the Anxiety Sensitivity Index. Self-reported relief and physiological tension were compared 5s before and after instructed deep breaths and breath holds, and before and after spontaneous deep breaths and breath holds in the respective blocks. Results show that self-reported relief following an instructed deep breath was higher than before. Physiological tension decreased following a spontaneous sigh in high anxiety sensitive persons and following a spontaneous breath hold in low anxiety sensitive persons. These results are the first to show that a deep breath relieves and, in anxiety sensitive persons, reduces physiological tension. These findings support the hypothesis that sighs are psychological and physiological resetters. Copyright © 2016 Elsevier Inc. All rights reserved.

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

Plant experience with check valves in passive systems

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Grush, W.H.; Koizumi, Y.; Woerth, S.C.

1983-01-01

An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data

Application of the ALARA principle to the bleed condenser relief valve replacement project

International Nuclear Information System (INIS)

Wong, T.; Kraft, J.

1997-01-01

Darlington Nuclear Generating Division's achievements in radiation dose performance can be attributed, in part, to adherence to the ALARA principle. The station's ALARA program, which is an integral part of the safety culture, derives its strength from a strong and unwavering commitment by the site Vice President. This commitment is supported by performance standards and an accountability system which holds managers and supervisors responsible for dose performance. A LAN-based hazard and dose information system with site-wide accessibility was established to facilitate work planning and exposure control. The principle of dose optimization is fully integrated into the work management process and includes work planning, progress monitoring and post-job review. An integrated performance assessment and reporting system also provides timely feedback to management on dose performance. An example of the ALARA program was the bleed condenser relieve valve replacement project. Pre-job ALARA review meetings for the project were held with supervisory staff and technicians to discuss job details and dose reduction measures. All work groups were required to prepare a detailed step be step task safety analysis (TSA). The trades and engineering staff were requested to critique the TSA and suggest ways of reducing dose. Over 30 practical ALARA suggestions were received and adopted for implementation. Daily meetings were held to review job progress and the effectiveness exposure control. A post-job ALARA review was held at the conclusion of each project to obtain feedback and lessons learned. All improvement suggestions were reviewed for implementation during subsequent installations. As a result of the ALARA initiatives, significant dose savings were achieved. The normalized dose expenditure has been reduced from 9.6 rem for the first installation to 6.9 rem for the fourth and last installation. (author)

Safety Evaluation Report, pump and valve inservice testing program, Yankee Rowe Nuclear Power Station (Docket No. 50-29). Revision 1

International Nuclear Information System (INIS)

Ransom, C.B.; Rockhold, H.C.

1985-10-01

The staff reviewed the Yankee Rowe IST program submittal dated October 2, 1981 and evaluated the proposed tests for compliance with the requirements of the ASME Boiler and Pressure Vessel Code, Section XI, 1977 Edition, through the Summer of 1978 Addenda. Those items not in compliance were discussed in a working session with Yankee Atomic Electric Company personnel, USNRC representatives, and EG and G Idaho reviewers on June 22, 1982. The evaluations presented in this SER of the Yankee Rowe pump and valve inservice testing program and the associated relief requests are those of the NRC staff. These findings apply only to component testing

Platypnea-Orthodeoxia Syndrome after Transcatheter Aortic Valve Implantation

Directory of Open Access Journals (Sweden)

Andrew K. Roy

2016-01-01

Full Text Available Progressive dyspnea and hypoxaemia in the subacute phase after transcatheter aortic valve implantation (TAVI are uncommon and warrant immediate assessment of valve and prosthesis leaflet function to exclude thrombosis, as well as investigation for other causes related to the procedure, such as left ventricular dysfunction, pulmonary embolism, and respiratory sepsis. In this case, we report the observation of a patient presenting two weeks after TAVI with arterial hypoxaemia in an upright position, relieved by lying flat, and coupled with an intracardiac shunt detected on echocardiography in the absence of pulmonary hypertension, raising the suspicion of Platypnea-Orthodeoxia Syndrome (POS. Invasive intracardiac haemodynamic assessment showed a significant right-to-left shunt (Qp/Qs = 0.74, which confirmed the diagnosis, with subsequent closure of the intracardiac defect resulting in immediate relief of symptoms and hypoxaemia. To our knowledge, this is the first reported case of an interatrial defect and shunt causing Platypnea-Orthodeoxia Syndrome after transcatheter aortic valve implantation, resolved by percutaneous device closure.

Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants

International Nuclear Information System (INIS)

Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

1995-05-01

Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures

Efficacy and safety of transcatheter aortic valve replacement in aortic stenosis patients at low to moderate surgical risk: a comprehensive meta-analysis

OpenAIRE

Elmaraezy, Ahmed; Ismail, Ammar; Abushouk, Abdelrahman Ibrahim; Eltoomy, Moutaz; Saad, Soha; Negida, Ahmed; Abdelaty, Osama Mahmoud; Abdallah, Ahmed Ramadan; Aboelfotoh, Ahmed Magdy; Hassan, Hossam Mahmoud; Elmaraezy, Aya Gamal; Morsi, Mahmoud; Althaher, Farah; Althaher, Moath; AlSafadi, Ammar M.

2017-01-01

Background Recently, transcatheter aortic valve replacement (TAVR) has become the procedure of choice in high surgical risk patients with aortic stenosis (AS). However, its value is still debated in operable AS cases. We performed this meta-analysis to compare the safety and efficacy of TAVR to surgical aortic valve replacement (SAVR) in low-to-moderate surgical risk patients with AS. Methods A systematic search of five authentic databases retrieved 11 eligible studies (20,056 patients). Rele...

Transcatheter Mitral Valve Devices - Functional Mechanical Designs.

Science.gov (United States)

Kliger, Chad

2014-03-01

Mitral regurgitation is a complex disorder involving a multitude of components of the mitral apparatus. With the desire for less invasive treatment approaches, transcatheter mitral valve therapies (TMVT) are directed at these components and available at varying stages of development. Therapeutic advancements and the potential to combine technologies may further improve their efficacy and safety. Transcatheter mitral valve replacement, while preserving the mitral apparatus, may emerge as an alternative or even a more suitable treatment option. In addition, early data on transcatheter mitral valve-in-valve and valve-in-ring implantation are encouraging and this approach may be an alternative to reoperation in the high-risk patient. This review details the expanding functional mechanical designs of current active TMVT.

Feasibility study for the adoption of POSRV for KNGR safety depressurization system

International Nuclear Information System (INIS)

Kwon, Young Min; Lim, Hong Sik; Song, Jin Ho; Sim, Suk Ku; Park, Jong Kyun

1999-03-01

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature of safety depressurization system(SDS) to rapidly de pressurize the reactor coolant system(RCS) in case of beyond design basis events of severe accidents, or a highly unlikely event of a total loss of feedwater (TLOFW) to both steam generators. Two design approaches were considered for the KNGR SDS design. The use of bleed valves similar to those of ABB-CE's system 80+ is design option 1, while in design option 2, the Power Operated Safety Relief valve (POSRV) is considered to provide the combined function of overpressure protection and rapid depressurization. The purpose of this report is to investigate the feasibility of adoption of French SebimPOSRVs for KNGR SDS (design option 2). This report provides the methodology to analyze the TLOFW event with Sebim valves and presents the results of thermal hydraulic analyses using a best-estimate version CEFLASH-4AS/REM for the TLOFW event with feed and bleed. The analyses were performed using a preliminary KNGR design data. For design option 2, if the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible to mitigate the consequences of the TLOFW event with a sufficient margin. (Author). 22 refs., 3 tabs., 19 figs

Improvement of a feed and bleed process for a 900 MWe NPP using the SIMPACT simulator

International Nuclear Information System (INIS)

Pochard, R.; Jedrzejewski, F.; Mazauric, X.; Cartenon, P. Y.

2000-01-01

A sensitivity study related to the improvement of a feed and bleed process was carried out with the SIPACT simulator. The scenario analysed here is related to a total loss of feed water on a French 900MW NPP. In a previous study we were looking extensively to the effects of bleeding with the three relief valves and its time of initiation. In the new calculations, limited bleed, by opening only one or two relief valves, was initiated at the minimum of mass and at the time which corresponds to the beginning of the heat transfer degradation in the steam generators. The analysis of the results shows that the in-vessel mass and the safety were improved when the number of actuated relief valves was reduced. But on the contrary the pressure reduction was limited by the performance of the HPIS and the equilibrium with the outlet flow from the relief valves. From these results a scenario with the consecutive opening of the three relief valves so as to depressurise while at the same time trying to optimise the in-vessel mass balance was proposed with a possible automation. (author)

The JUPITER registry: One-year outcomes of transapical aortic valve implantation using a second generation transcatheter heart valve for aortic regurgitation.

Science.gov (United States)

Silaschi, Miriam; Conradi, Lenard; Wendler, Olaf; Schlingloff, Friederike; Kappert, Utz; Rastan, Ardawan J; Baumbach, Hardy; Holzhey, David; Eichinger, Walter; Bader, Ralf; Treede, Hendrik

2018-06-01

We present 1-year outcomes of the post-market registry of a next-generation transcatheter heart valve used for aortic regurgitation (AR). Transcatheter aortic valve replacement (TAVR) is routine in high-risk patients with aortic stenosis but is not recommended for AR. The JenaValve™ (JenaValve Technology GmbH, Munich, Germany) overcomes technical challenges in AR patients through a leaflet clipping mechanism. The JenaValve EvalUation of Long Term Performance and Safety In PaTients with SEvere Aortic Stenosis oR Aortic Insufficiency (JUPITER) Registry is a European study to evaluate safety and effectiveness of this THV. From 2012-2015, 30 patients with AR were enrolled. Mean age was 74.4 ± 9.3 years. Procedural success was 96.7% (29/30). One patient was converted to open surgery. No annular rupture or coronary ostia obstruction occurred. Mortality at 30 days was 10.0% (3/30). Combined safety endpoint was met in 13.3% (4/30). Paravalvular regurgitation was not present/trivial in 84.6% (22/26) and mild in 15.4% (4/26). Rate of permanent pacemaker implantation was 3.8% (1/26). One-year Kaplan-Meier survival was 79.9%, one-year combined efficacy was 73.1% (19/30). No further strokes were observed during 1 year of follow-up. The JenaValve overcomes technical challenges of TAVR in AR through a clipping mechanism. We report satisfactory outcomes of a multicenter registry using the JenaValve for predominant AR, as rate of THV embolization, residual AR and permanent pacemaker implantation was low. One-year results using the JenaValve for AR encourage its use for this indication. © 2017 Wiley Periodicals, Inc.

Function analysis of steam isolation valves

International Nuclear Information System (INIS)

Persson, R.; Sjoeberg, A.; Aakesson, H.; Kilpi, K.; Noro, H.; Siikonen, T.; Wallen, G.

1981-01-01

Function analysis of system-medium-operated steam isolation valves has been the objective of the Swedish-Finnish IVLS project, the results of which are presented in this report. Theoretical models were to be verified against available experimental data, to some extent from the HDR blowdown experiments, which are part of a German reactor safety program. Finnish hydraulic measurements on a valve model (scale 1:2.15) have been performed to give complementary data. The analysis work has covered the thermal-hydraulic behaviour of steam isolation valves as well as phenomena related to structural mechanics. (Auth.)

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.

Efficacy and safety of intravitreal bevacizumab in eyes with neovascular glaucoma undergoing Ahmed glaucoma valve implantation: 2-year follow-up.

Science.gov (United States)

Arcieri, Enyr S; Paula, Jayter S; Jorge, Rodrigo; Barella, Kleyton A; Arcieri, Rafael S; Secches, Danilo J; Costa, Vital P

2015-02-01

To evaluate the efficacy and safety of intravitreal bevacizumab (IVB) in eyes with neovascular glaucoma (NVG) undergoing Ahmed glaucoma valve (AGV) implantation. This was a multicentre, prospective, randomized clinical trial that enrolled 40 patients with uncontrolled neovascular glaucoma that had undergone panretinal photocoagulation and required glaucoma drainage device implantation. Patients were randomized to receive IVB (1.25 mg) or not during Ahmed valve implant surgery. Injections were administered intra-operatively, and 4 and 8 weeks after surgery. After a mean follow-up of 2.25 ± 0.67 years (range 1.5-3 years), both groups showed a significant decrease in IOP (p glaucoma undergoing Ahmed glaucoma valve implantation. There is a trend to slightly lower IOPs and number of medications with IVB use during AGV implantation for neovascular glaucoma. © 2014 Acta Ophthalmologica Scandinavica Foundation. Published by John Wiley & Sons Ltd.

Gate valve and motor-operator research findings

International Nuclear Information System (INIS)

Steele, R. Jr.; DeWall, K.G.; Watkins, J.C.; Russell, M.J.; Bramwell, D.

1995-09-01

This report provides an update on the valve research being sponsored by the US Nuclear Regulatory Commission (NRC) and conducted at the Idaho National Engineering Laboratory (INEL). The research addresses the need to provide assurance that motor-operated valves can perform their intended safety function, usually to open or close against specified (design basis) flow and pressure loads. This report describes several important developments: Two methods for estimating or bounding the design basis stem factor (in rising-stem valves), using data from tests less severe than design basis tests; a new correlation for evaluating the opening responses of gate valves and for predicting opening requirements; an extrapolation method that uses the results of a best effort flow test to estimate the design basis closing requirements of a gate valve that exhibits atypical responses (peak force occurs before flow isolation); and the extension of the original INEL closing correlation to include low- flow and low-pressure loads. The report also includes a general approach, presented in step-by-step format, for determining operating margins for rising-stem valves (gate valves and globe valves) as well as quarter-turn valves (ball valves and butterfly valves)

The 1994 loss of coolant incident at Pickering NGS

Energy Technology Data Exchange (ETDEWEB)

Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

Initial data collection efforts of CREDO. Sodium valve failures

International Nuclear Information System (INIS)

Bott, T.F.; Haas, P.M.

1978-01-01

The Centralized Reliability Data organisation (CREDO) has been established at Oak Ridge National Laboratory to define, develop, and maintain a reliability data analysis center for use in advanced reactor safety and licensing. Its primary functions are collection, reduction, evaluation, storage, retrieval, and dissemination of reliability/maintainability data. Data-collection efforts have been initiated at several test loops, at the Experimental Breeder Reactor-II and at the Fast Flux Test Facility. Top priority is being given to collection data on safety and safety-related systems, primarily for sodium-cooled reactors. Sufficient operating time has been accumulated on sodium valves at test facilities to provide quantitative estimates of reliability characteristics with a reasonable degree of confidence. Sodium-valve failures have been categorized according to seat design, size, seal type, and actuator type. Attempts have been made to establish the variation of failure rate with time and duty. Estimates of failure rates for sodium valves have been compared to those for water valves and appear to be of the same order of magnitude. (author)

Influence of the CVCS Modelling on Results of the Loss of Offsite Power (LOOP) Safety Analysis for NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Bajs, T.; Debrecin, N.

2006-01-01

A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief

A strategy for the risk-based inspection of pressure safety valves

International Nuclear Information System (INIS)

Chien, C.-H.; Chen, C.-H.; Chao, Y.J.

2009-01-01

The purpose of a pressure safety valve (PSV) is to protect the life and safety of pressure vessels in a pressurized system. If a weakened PSV fails to function properly, a catastrophic event might occur if no other protective means are provided. By utilizing the as-received test data and statistical analysis of the aging conditions of PSVs in lubricant process units, a risk-based inspection (RBI) system was developed in this study. First of all, the characteristics of PSV were discussed from the practical viewpoint of engineering inspection and maintenance. The as-received test data, which shows obvious PSV damage, will be separated from the data used in the following statistical analysis. Then, the relationship between the aging conditions and the corresponding PSV parameters was analyzed by using the statistical technique-analysis of variance (ANOVA). Finally, a strategy for semi-quantitative RBI is proposed. Also, a definitive estimated inspection interval for every PSV is suggested. The outcome indicated most of the risks result from a few PSVs, for which the corresponding inspection intervals will be shorter than the 2 years in accordance with relative standards and local government regulations

ADAM®/SIPLUG®: An innovative valve monitoring system

International Nuclear Information System (INIS)

Muñoz, L.; Krell, M.

2012-01-01

Optimized maintenance strategies are a key aspect for safe and undisturbed plant operation. Innovative valve service solutions, e.g. valve diagnostics can support this in an efficient way. The ADAM®/SIPLUG® valve monitoring system allows full online monitoring of valves and actuators with automatic evaluation and assessment. Especially for safety-related and operation-related valves this provides valuable information on components condition to ensure proper function and contribute to optimization of maintenance strategies as well as effective maintenance performance. The new SIPLUG®-4 modules are the evolutionary solution for valve diagnosis at the Motor Control Center (MCC). As the SIPLUG®-4 can be installed directly in the MCC outgoing actuator power cable it allows an easy installation in existing switchgear cabinets. Measurement at MCC means also zero effort for performance of diagnostics reducing the number of on-site activities. This results in decrease of maintenance costs and dose rates for deployed personnel. The ADAM® evaluation software and database was developed in parallel with the hardware. It provides automatic analysis of the monitoring results using the limit values specified for the valves. The measured data can be transmitted via the power plant’s local area network to the ADAM® data server, if the SIPLUG® online hardware is installed. With the mobile solution, the data can be transmitted via serial or USB interface to a PC or notebook. With this solution all measurement information will be available immediately in the offices of plant engineers. Also, with SIPLUG® online all operations of valves can be automatically recorded. More than 25 years of experience in various plants worldwide show that the application of ADAM®/SIPLUG® valve diagnostics solution leads to increased plant safety and availability. Some of the references for ADAM®/SIPLUG® are Germany, Switzerland, Brazil, Spain and Eastern Europe. (author)

46 CFR 151.50-31 - Chlorine.

Science.gov (United States)

2010-10-01

... inside diameter manhole, fitted with a cover located above the maximum liquid level and as close as... accessary equipment shall be of a type suitable for use with chlorine and shall be made of metal, corrosion... shutoff valves and with safety relief valves. All valves shall be bolted to the cover or covers specified...

Structural Safety Analysis Based on Seismic Service Conditions for Butterfly Valves in a Nuclear Power Plant

Directory of Open Access Journals (Sweden)

Sang-Uk Han

2014-01-01

Full Text Available The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements.

Colombian experience with transcatheter aortic valve implantation of medtronic CoreValve.

Science.gov (United States)

Dager, Antonio E; Nuis, Rutger-Jan; Caicedo, Bernardo; Fonseca, Jaime A; Arana, Camilo; Cruz, Lidsa; Benitez, Luis M; Nader, Carlos A; Duenas, Eduardo; de Marchena, Eduardo J; O'Neill, William W; de Jaegere, Peter P

2012-01-01

At our institutions, increasing numbers of aortic stenosis patients were not candidates for surgical aortic valve replacement. Accordingly, we initiated the Cali Colombian Transcatheter Aortic Valve Implantation (TAVI) program. From March 2008 through January 2011, 53 consecutive patients (mean age, 79 ± 6 yr; men, 58%) underwent TAVI with the Medtronic CoreValve System, and data were prospectively collected. Our study's endpoints conformed with Valve Academic Research Consortium recommendations. We report our clinical results.Predicted mortality rates were 25% (interquartile range, 17%-34%) according to logistic EuroSCORE and 6% (interquartile range, 3%-8%) according to the Society of Thoracic Surgeons score. The 30-day mortality rate was 9% (3 intraprocedural deaths, 5 total). The combined 30-day safety endpoint was 30% (major vascular sequelae, 23%; life-threatening bleeding, 12%; myocardial infarction, 4%; major stroke, 4%; and acute kidney injury [stage 3], 2%). Eight patients (15%) required post-implantation balloon dilation and 2 (4%) required valve-in-valve implantation, for a technical device success rate of 77%. Mean peak transvalvular gradient decreased from 74 ± 29 to 17 ± 8 mmHg and mean transvalvular gradient from 40 ± 17 to 8 ± 4 mmHg (both P=0.001). Moderate or severe aortic regurgitation decreased from 32% to 18% (P=0.12) and mitral regurgitation from 32% to 13% (P=0.002). The 1-year survival rate was 81%.We found that TAVI with the CoreValve prosthesis was safe and feasible, with sustained long-term results, for treating aortic stenosis in patients at excessive surgical risk; nonetheless, serious adverse events occurred in 30% of the patients.

Evaluation of operating experience with safety values

International Nuclear Information System (INIS)

Bung, W.; Hoemke, P.; Oberender, W.; Paul, H.; Rueter, W.

1985-01-01

This report describes statistical investigations of 2076 functional tests carried out on power operated safety valves in conventional power plants in 1972 until 1983 with special regard to Common Mode-Failures. The results clearly show that Common Mode-Failures play an important part of non-availability for the controlled safety valves, especially in the control system. The 'Deutsche Risikostudie' does not consider any Common Mode-Failures of the primary safety valves. However there is no significant increase of the risk resulted by the primary safety valves in the 'Referenzanlage' if the calculated Common Mode-Failures probabilities are considered. (orig.) [de

Basis of valve operator selection for SMART

International Nuclear Information System (INIS)

Kang, H. S.; Lee, D. J.; See, J. K.; Park, C. K.; Choi, B. S.

2000-05-01

SMART, an integral reactor with enhanced safety and operability, is under development for use of the nuclear energy. The valve operator of SMART system were selected through the data survey and technical review of potential valve fabrication vendors, and it will provide the establishment and optimization of the basic system design of SMART. In order to establish and optimize the basic system design of SMART, the basis of selection for the valve operator type were provided based on the basic design requirements. The basis of valve operator selection for SMART will be used as a basic technical data for the SMART basic and detail design and a fundamental material for the new reactor development in the future

Basis of valve operator selection for SMART

Energy Technology Data Exchange (ETDEWEB)

Kang, H. S.; Lee, D. J.; See, J. K.; Park, C. K.; Choi, B. S

2000-05-01

SMART, an integral reactor with enhanced safety and operability, is under development for use of the nuclear energy. The valve operator of SMART system were selected through the data survey and technical review of potential valve fabrication vendors, and it will provide the establishment and optimization of the basic system design of SMART. In order to establish and optimize the basic system design of SMART, the basis of selection for the valve operator type were provided based on the basic design requirements. The basis of valve operator selection for SMART will be used as a basic technical data for the SMART basic and detail design and a fundamental material for the new reactor development in the future.

CONSIDERING TRAVEL TIME RELIABILITY AND SAFETY FOR EVALUATION OF CONGESTION RELIEF SCHEMES ON EXPRESSWAY SEGMENTS

Directory of Open Access Journals (Sweden)

Babak MEHRAN

2009-01-01

Full Text Available Evaluation of the efficiency of congestion relief schemes on expressways has generally been based on average travel time analysis. However, road authorities are much more interested in knowing the possible impacts of improvement schemes on safety and travel time reliability prior to implementing them in real conditions. A methodology is presented to estimate travel time reliability based on modeling travel time variations as a function of demand, capacity and weather conditions. For a subject expressway segment, patterns of demand and capacity were generated for each 5-minute interval over a year by using the Monte-Carlo simulation technique, and accidents were generated randomly according to traffic conditions. A whole year analysis was performed by comparing demand and available capacity for each scenario and shockwave analysis was used to estimate the queue length at each time interval. Travel times were estimated from refined speed-flow relationships and buffer time index was estimated as a measure of travel time reliability. it was shown that the estimated reliability measures and predicted number of accidents are very close to observed values through empirical data. After validation, the methodology was applied to assess the impact of two alternative congestion relief schemes on a subject expressway segment. one alternative was to open the hard shoulder to traffic during the peak period, while the other was to reduce the peak period demand by 15%. The extent of improvements in travel conditions and safety, likewise the reduction in road users' costs after implementing each improvement scheme were estimated. it was shown that both strategies can result in up to 23% reduction in the number of occurred accidents and significant improvements in travel time reliability. Finally, the advantages and challenging issues of selecting each improvement scheme were discussed.

Function analysis of steam isolation valves

International Nuclear Information System (INIS)

Persson, R.; Kilpi, K.; Noro, H.; Siikonen, T.; Sjoeberg, A.; Wallen, G.; Aakesson, H.

1981-01-01

Function analysis of system-medium-operated steam isolation valves has been the objective of the Swedish-Finnish IVLS project, the results of which are presented in this report. Theoretical models were to be verified against available experimental data, to some extent from the HDR blowdown experiments, which are part of a German reactor safety program. Finnish hydraulic measurements on a valve model (scale 1:2.15) have been performed to give complementary data. The analysis work has covered the thermal-hydraulic behaviour of steam isolation valves as well as phenomena related to structural mechanics. Work performed under contract with the Swedish Nuclear Power Inspectorate. (Author)

Flowfield Analysis of a Pneumatic Solenoid Valve

Directory of Open Access Journals (Sweden)

Sheam-Chyun Lin

2018-07-01

Full Text Available Pneumatic solenoid valve has been widely used in the vehicle control systems for meeting the rapid-reaction demand triggered by the dynamic conditions encountered during the driving course of vehicle. For ensuring the safety of human being, the reliable and effective solenoid valve is in great demand to shorten the reaction time and thus becomes the topic of this research. This numerical study chooses a commercial 3/2-way solenoid valve as the reference valve for analysing its performance. At first, CFD software Fluent is adopted to simulate the flow field associated with the valve configuration. Then, the comprehensive flow visualization is implemented to identify the locations of adverse flow patterns. Accordingly, it is found that a high-pressure region exists in the zone between the nozzle exit and the top of iron core. Thereafter, the nozzle diameter and the distance between nozzle and spool are identified as the important design parameters for improving the pressure response characteristics of valve. In conclusion, this work establishes a rigorous and systematic CFD scheme to evaluate the performance of pneumatic solenoid valve.

Analysis of an ADS spurious opening event at a BWR/6 by means of the TRACE code

International Nuclear Information System (INIS)

Nikitin, Konstantin; Manera, Annalisa

2011-01-01

Highlights: → The spurious opening of 8 relief valves of the ADS system in a BWR/6 has been simulated. → The valves opening results in a fast depressurization and significant loads on the RPV internals. → This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the available plant data. - Abstract: The paper presents the results of a post-event analysis of a spurious opening of 8 relief valves of the automatic depressurization system (ADS) occurred in a BWR/6. The opening of the relief valves results in a fast depressurization (pressure blow down) of the primary system which might lead to significant dynamic loads on the RPV and associated internals. In addition, the RPV level swelling caused by the fast depressurization might lead to undesired water carry-over into the steam line and through the safety relief valves (SRVs). Therefore, the transient needs to be characterized in terms of evolution of pressure, temperature and fluid distribution in the system. This event has been modeled by means of the TRACE and TRAC-BF1 codes. The results are in good agreement with the plant data.

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.)

Aortic Regurgitation in Patients Undergoing Transcatheter Aortic Valve Replacement With the Self-Expanding CoreValve Versus the Balloon-Expandable SAPIEN XT Valve.

Science.gov (United States)

Kiramijyan, Sarkis; Magalhaes, Marco A; Koifman, Edward; Didier, Romain; Escarcega, Ricardo O; Baker, Nevin C; Negi, Smita I; Minha, Sa'ar; Torguson, Rebecca; Jiaxiang, Gai; Asch, Federico M; Wang, Zuyue; Okubagzi, Petros; Gaglia, Michael A; Ben-Dor, Itsik; Satler, Lowell F; Pichard, Augusto D; Waksman, Ron

2016-05-01

The incidence of aortic regurgitation (AR) after transcatheter aortic valve replacement (TAVR) in a self-expanding and a balloon-expandable system is controversial. This study aimed to examine the incidence and severity of post-TAVR AR with the CoreValve (CV) versus the Edwards XT Valve (XT). Baseline, procedural, and postprocedural inhospital outcomes were compared. The primary end point was the incidence of post-TAVR AR of any severity, assessed with a transthoracic echocardiogram, in the CV versus XT groups. A multivariate logistic regression analysis was completed to evaluate for correlates of the primary end point. The secondary end points included the change in severity of AR at 30-day and 1-year follow-up. A total of 223 consecutive patients (53% men, mean age 82 years) who had transfemoral TAVR with either a CV (n = 119) or XT (n = 104) were evaluated. The rates of post-TAVR AR in the groups were similar, and there was no evidence of more-than-moderate AR in either group. There were significant differences in the rates of intraprocedural balloon postdilation with the CV (17.1%) versus XT valve (5.8%; p = 0.009) and in the rates of intraprocedural implantation of a second valve-in-valve prosthesis with the CV (9.9%) versus XT valve (2.2%; p = 0.036). There were no significant differences in inhospital safety outcomes between the 2 groups. In conclusion, the incidence of post-TAVR AR is similar between the CV and the XT valve when performed by experienced operators using optimal intraprocedural strategies, as deemed appropriate, to mitigate the severity of AR. Copyright © 2016 Elsevier Inc. All rights reserved.

SOLA-LOOP analysis of a back pressure check valve

International Nuclear Information System (INIS)

Travis, J.R.

1984-01-01

The SOLA-LOOP computer code for transient, nonequilibrium, two-phase flows in networks has been coupled with a simple valve model to analyze a feedwater pipe breakage with a back-pressure check valve. Three tests from the Superheated Steam Reactor Safety Program Project (PHDR) at Kahl, West Germany, are analyzed, and the calculated transient back-pressure check valve behavior and fluid dynamics effects are found to be in excellent agreement with the experimentally measured data

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

Pressurizer safety valve serviceability enhancement by spring compression stability

Energy Technology Data Exchange (ETDEWEB)

Ratiu, M.D.; Moisidis, N.T. [California Consulting Engineering and Technology (CALCET), San Leandro, California (United States)

2007-07-01

The proactive maintenance of the spring-loaded-self-actuated Pressurizer Safety Valve (PSV) has caused frequent concerns pertaining the spring self actuated reliability due to set point drift, spurious openings, and seat leakage. The exhaustive testing performed on a Crosby PSV model 6M6 has revealed that the principal cause of these malfunctions is the spring compression elastic instability during service. The spring lateral deformations measurements performed validated the analytical shapes for spring compression: symmetrical bending - for coaxial supported ends - restraining any support displacement, and asymmetrical bending induced by the potential misalignment of the supported top end. The source of the spring compression instability appears on the tested Crosby PSV induced by the top end lateral displacement during long term operation. The testing with restrained displacement at the spring top has shown consistent set-point reproducibility, less than +/- 1 per cent. To eliminate the asymmetrical spring buckling, a design review of the PSV is proposed including the guided fixture at the top and the decrease of spring coil slenderness ratio H/D, corresponding to the general analytical elastic stability for the asymmetrical compression. (authors)

Pressurizer safety valve serviceability enhancement by spring compression stability

International Nuclear Information System (INIS)

Ratiu, M.D.; Moisidis, N.T.

2007-01-01

The proactive maintenance of the spring-loaded-self-actuated Pressurizer Safety Valve (PSV) has caused frequent concerns pertaining the spring self actuated reliability due to set point drift, spurious openings, and seat leakage. The exhaustive testing performed on a Crosby PSV model 6M6 has revealed that the principal cause of these malfunctions is the spring compression elastic instability during service. The spring lateral deformations measurements performed validated the analytical shapes for spring compression: symmetrical bending - for coaxial supported ends - restraining any support displacement, and asymmetrical bending induced by the potential misalignment of the supported top end. The source of the spring compression instability appears on the tested Crosby PSV induced by the top end lateral displacement during long term operation. The testing with restrained displacement at the spring top has shown consistent set-point reproducibility, less than +/- 1 per cent. To eliminate the asymmetrical spring buckling, a design review of the PSV is proposed including the guided fixture at the top and the decrease of spring coil slenderness ratio H/D, corresponding to the general analytical elastic stability for the asymmetrical compression. (authors)

The use of transcatheter aortic valve replacement vs surgical aortic valve replacement for the treatment of aortic stenosis

Directory of Open Access Journals (Sweden)

Jensen HA

2015-08-01

Full Text Available Hanna A Jensen, Lillian L Tsai, Vinod H Thourani Division of Cardiothoracic Surgery, Joseph B Whitehead Department of Surgery, Structural Heart and Valve Center, Emory University School of Medicine, Atlanta, GA, USA Abstract: Severe aortic stenosis (AS is associated with considerable morbidity and mortality and is increasing in prevalence as the global population increases. Since AS primarily affects the elderly, many of these patients have comorbidities that make them poor candidates for the gold standard treatment for AS, surgical aortic valve replacement. Transcatheter aortic valve replacement has emerged as a novel technology for the management of AS in higher risk patients over the past decade. Randomized trials have established the safety and efficacy of transcatheter aortic valve replacement, and the medical community has rallied to identify the patients who are most suitable for this transformative treatment. This review focuses on outlining the key procedural differences, describing the unique challenges of both operations, and finally assessing and comparing outcomes both on a general level and in challenging patient subgroups. Keywords: aortic valve replacement, transcatheter aortic valve replacement, surgical aortic valve replacement 

A review of digital radiography technology for valve inspection

International Nuclear Information System (INIS)

Stoev, K.; Guerout, F.M.; Horn, D.

2008-01-01

There are thousands of valves in a nuclear power plant (NPP) used for control, safety and checks in various plant systems, so there is a well-identified need for fast and reliable inspection and diagnostics of valves. Digital radiography can provide considerable improvements to the inspection and testing procedures for valves in comparison to classical film radiography. These improvements can lead to significant financial advantages by providing real-time inspection results, significantly reduced inspection and decision-making time, and reduced operational cost. Digital image processing, including digital image enhancement, digital archiving, and digital communication of the images and the results, is also a considerable advantage over classical film radiography technology. Another advantage of digital radiography technology is the improved safety and the reduced environmental impact due to reduced exposure/test times, use of smaller exclusion zones, elimination of chemical processing, and absence of disposable materials. This paper reviews the existing technology and evaluates the potential of digital radiography for inspection and diagnostics of valves. Station needs and requirements are assessed, and the safety, environmental and economical constraints of digital radiography techniques estimated. The advantages and disadvantages of different digital radiography equipment are compared, and their limitations and characteristics studied. (author)

Theoretical seismic analysis of butterfly valve for nuclear power plant

International Nuclear Information System (INIS)

Han, Sang Uk; Ahn, Jun Tae; Han, Seung Ho; Lee, Kyung Chul

2012-01-01

Valves are one of the most important components of a pipeline system in a nuclear power plant, and it is important to ensure their structural safety under seismic loads. A crucial aspect of structural safety verification is the seismic qualification, and therefore, an optimal shape design and experimental seismic qualification is necessary in case the configuration of the valve parts needs to be modified and their performance needs to be improved. Recently, intensive numerical analyses have been preformed before the experimental verification in order to determine the appropriate design variables that satisfy the performance requirements under seismic loads. In this study, static and dynamic numerical structural analyses of a 200A butterfly valve for a nuclear power plant were performed according to the KEPIC MFA. The result of static analysis considering an equivalent static load under SSE condition gave an applied stress of 135MPa. In addition, the result of dynamic analysis gave an applied stress of 183MPa, where the CQC method using response spectrums was taken into account. These values are under the allowable strength of the materials used for manufacturing the butterfly valve, and therefore, its structural safety satisfies the requirements of KEPIC MFA

Theoretical seismic analysis of butterfly valve for nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Han, Sang Uk; Ahn, Jun Tae; Han, Seung Ho [Donga Univ., Busan (Korea, Republic of); Lee, Kyung Chul [Dukwon Valve Co., Ltd., Busan (Korea, Republic of)

2012-09-15

Valves are one of the most important components of a pipeline system in a nuclear power plant, and it is important to ensure their structural safety under seismic loads. A crucial aspect of structural safety verification is the seismic qualification, and therefore, an optimal shape design and experimental seismic qualification is necessary in case the configuration of the valve parts needs to be modified and their performance needs to be improved. Recently, intensive numerical analyses have been preformed before the experimental verification in order to determine the appropriate design variables that satisfy the performance requirements under seismic loads. In this study, static and dynamic numerical structural analyses of a 200A butterfly valve for a nuclear power plant were performed according to the KEPIC MFA. The result of static analysis considering an equivalent static load under SSE condition gave an applied stress of 135MPa. In addition, the result of dynamic analysis gave an applied stress of 183MPa, where the CQC method using response spectrums was taken into account. These values are under the allowable strength of the materials used for manufacturing the butterfly valve, and therefore, its structural safety satisfies the requirements of KEPIC MFA.

Independent deterministic analysis of the operational event with turbine valve closure and one atmospheric dump valve stuck open

International Nuclear Information System (INIS)

Rijova, N.

2007-01-01

The paper presents the results of the independent analysis of the operational event which took place on 07.11.2003 at Unit 1 of Rostov NPP. The event started with switching off the electrical generator of the turbine due to a short cut at the local switching substation. The turbine isolating valves closed to prevent damage of the turbine. The condenser dump valves (BRU-K) and the atmospheric dump valves (BRU-A) opened to release the vapour generated in the steam generators. After the pressure decrease in the steam generators BRU-K and BRU-A closed but one valve stuck opened. The emergency core cooling system was activated automatically. The main circulation pump of the loop corresponding to the steam generator with the stuck BRU-A was tripped. The stuck valve was closed by the operational stuff manually. No safety limits were violated. The analysis of the event was carried out using ATHLET code. A reasonable agreement was achieved between the calculated and measured values. (author)

Rotarex Group: diversified ranges; Rotarex Group: une offre diversifiee

Energy Technology Data Exchange (ETDEWEB)

Anon.

2000-09-01

The Rotarex Group, located at Lintgen (Luxembourg), is a multinational company chaired by Jean-Claude Schmitz. With a workforce of some 1,000 people in the world, this group is specialized in valves, accessories and pressure reducers for a wide range of gas applications: from LP Gas to special gas (VHP, cryogenics,...) as well for medical and industrial gas, and also for CNG. The group is present at the San Diego exhibition, mainly though its affiliate Rotarex North America, headed by Bert Pistor, vice-president of Ceodeux Inc. Rotarex North America is well known on the US market for its OPD device for propane cylinders. But the company will show also a new generation of valves for cylinders. The group will represent its different affiliates in the world (South America, Africa, Europe, Far East, Australia,...), as its expansion is widening: a subsidiary was launched last Spring in Morocco, a new plant is built in the Czech Republic. In Europe, Ceodeux LPG TEC is one of the main European manufacturers of valves and equipments for LPG and refrigerating gas. Having already reinforced its presence on the French and Italian markets, Ceodeux LPG TEC is expanding through technique on its traditional markets as well as on new markets. In the field of automotive LPG, the company offers series of equipments linked to the LPG tank: single devices or multi-valves. These are equipped with a safety relief valve and can be completed by a thermal safety system or a second safety relief valve. For France, all these systems are type approved according to UN Regulation 67-01. France is in implementing, one year ahead of other European countries, this new Regulation for new LP gas vehicles as well for vehicles called back for retrofitting with new safety valves. Some 100,000 vehicles will receive a 27 bar safety valve or a new multi-valve fitted with this safety valve. This retro-fitting operation will be paid for 50% by vehicle's owners and 50 % by French Authorities and LP Gas

Stop valve with automatic control and locking for nuclear reactors

International Nuclear Information System (INIS)

Chung, D.K.

1980-01-01

This invention generally concerns an automatic control and locking stop valve. Specifically it relates to the use of such a valve in a nuclear reactor of the type containing absorber elements supported by a fluid and intended for stopping the reactor in complete safety [fr

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

International Nuclear Information System (INIS)

Hackett, D.B.

1980-01-01

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

A knowledge-based assistant for valve maintenance planning

International Nuclear Information System (INIS)

Winter, M.J.; Danofsky, R.A.; Spinrad, B.I.; Howard, K.

1987-01-01

A knowledge-based program is being developed to assist engineers in maintenance planning for safety related, motor-operated valves at a boiling water reactor. The purpose of this project is to develop the general framework for a prototype system that demonstrates the capabilities for diagnosing valve symptoms and prescribing corrective maintenance, completing a portion of the Corrective Maintenance Action Request (CMAR) form which must be prepared for each job, and managing an interactive valve data base. Minimizing user input and providing output in a form that is familiar to the maintenance planning engineer are important goals for the program. This paper describes the present features of the valve maintenance advisory system which is currently being tested

Ultrasonic evaluation of heat treatment for stress relief in steel

International Nuclear Information System (INIS)

Bittencourt, Marcelo de S.Q.; Lamy, Carlos A.; Goncalves Filho, Orlando J.A.; Payao Filho, Joao da C.

2000-01-01

Residual stresses in materials arise due to the manufacturing processes. As a consequence, in the nuclear area some components must suffer a stress relief treatment according to strict criteria. Although these treatments are carefully carried on, concern with nuclear safety is constantly growing. This work proposes a nondestructive ultrasonic method to guarantee the efficiency of the heat treatment. It was used a short peened steel plate with tensile and compressive stresses which was submitted to a stress relief treatment. The results show that the proposed ultrasonic method could be used to confirm the efficiency of the stress relief heat treatment. (author)

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.

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)

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.)

BIF butterfly valve life extension at WNP-2

International Nuclear Information System (INIS)

Armstrong, D.

1991-01-01

Primary containment purging, venting, inerting, and reactor building ventilation at the WNP-2 plant are accomplished with a series of large butterfly valves. A total of 31 valves which are similar in design, but of different sizes, employ an elastomer to achieve sealing integrity when closed. These valves, which were originally manufactured by BIF, a unit of General Signal, range in size from 18 to 84 inches in diameter. Service life in the plant was much less than desired for safety-related equipment, and several seal failures had been experienced shortly after valve overhaul. This program covers a design change made to enhance performance of the elastomer seal to achieve a very meaningful life extension. While numerous configurations of BIF valves exist, this work relates only to the model 657 unit assembled with an elastomer seal mounted onto the valve disc by a stainless steel clamping ring held with studs and nuts. The problems encountered, and the steps taken to resolve the deficiencies may, however, be applicable to other butterfly valve configurations

Improvement of seawater booster pump outlet check valve

International Nuclear Information System (INIS)

Li Xuning; Du Yansong; Huang Huimin

2010-01-01

Conventional island seawater booster pump set of QNPC 310 MWe unit are very important in the whole circulating cooling system, and the integrate function of seawater booster pump outlet check valve is the foundation of steady operation of the seawater booster pump set. The article mainly introduce that through the analyses to the reason to the problem that the seawater booster pump outlet check valve of QNPC 310 MWe unit appeared in past years by our team, and considering the influence of operation condition and circumstance, the team improve the seawater booster pump outlet check valve from swing check valve to shuttle check valve which operate more appropriately in the system. By the test of continuous practice, we make further modification to the inner structure of shuttle check valve contrapuntally, and therefore we solve the problem in seawater booster pump outlet check valve fundamentally which has troubled the security of system operation in past years, so we realize the aim of technical improvement and ensure that the system operate in safety and stability. (authors)

Strength and water-tightness of the closure head and valves of a model cask under high external pressure

International Nuclear Information System (INIS)

Terada, O.; Kumada, M.; Hayakawa, T.; Mochizuki, S.; Ohrui, K.

1978-01-01

This paper describes experimental research on the strength and water-tightness of the closure head and attached valves of a model cask under high external pressure, in simulation of its having been accidentally lost in the deep sea. Both the external pressure tests and the corrosion tests were carried out using scale models of the closure head of an 80-ton spent-fuel shipping cask, and the full size pressure relief valves and drain valves which were to be attached to the actual cask. Based on the results of the above tests, evaluations were made, and new information was obtained on the pressure-proof strength and water-tightness of the closure head of the cask and the valves. Lastly, research which is being carried on in Japan on the pressure equalizer is also introduced

Operating experience feedback report - Solenoid-operated valve problems

International Nuclear Information System (INIS)

Ornstein, H.L.

1991-02-01

This report highlights significant operating events involving observed or potential common-mode failures of solenoid-operated valves (SOVs) in US plants. These events resulted in degradation or malfunction of multiple trains of safety systems as well as of multiple safety systems. On the basis of the evaluation of these events, the Office for Analysis and Evaluation of Operational Data (AEOD) of the US Nuclear Regulatory Commission (NRC) concludes that the problems with solenoid-operated valves are an important issue that needs additional NRC and industry attention. This report also provides AEOD's recommendations for actions to reduce the occurrence of SOV common-mode failures. 115 refs., 7 figs., 2 tabs

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

International Nuclear Information System (INIS)

Meseth, J.

2002-01-01

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

Improvements in valve reliability due to implementation of effective condition monitoring programs

International Nuclear Information System (INIS)

Hale, Stan

2003-01-01

Modern diagnostic systems for motor-operated valves, pneumatic control valves and checkvalves have facilitated a shift in the maintenance philosophy for valves and actuators in nuclear power plants from schedule based to condition-based maintenance (CBM). This shift enables plant management to focus resources and schedule priority on the plant equipment that warrants attention thereby not wasting resources or increasing the human factors risk on equipment that has not degraded. The most recent initiatives combine condition monitoring with risk/safety insights to focus attention and resonances on the right equipment at the right time consistent with each component's safety-significance. The activities of the ASME working groups responsible for nuclear O and M codes have kept pace with the technology and process improvements necessary to maximize the technical and economic benefits of condition based and risk informed maintenance. This paper discusses adoption of valve condition monitoring in the nuclear power industry, changes to ASME codes and standards during the 90's to facilitate adoption of condition monitoring technology for in-service testing and recent efforts to combine risk insights with condition monitoring strategies to achieve the highest level of valve reliability and nuclear safety without over inflating maintenance cost. (author)

Analysis of tank safety with propane-butane on LPG distribution station

Directory of Open Access Journals (Sweden)

Krzysiak Zbigniew

2017-12-01

Full Text Available An analysis of the risk of failure in the safety valve – tank with propane-butane (LPG system has been conducted. An uncontrolled outflow of liquid LPG, caused by a failure of the above mentioned system has been considered as a threat. The main research goal of the study is the hazardous analysis of propane-butane gas outflow for the safety valve – LPG tank system. The additional goal is the development of an useful method to fast identify the hazard of a mismatched safety valve. The results of the research analysis have confirmed that safety valves are basic protection of the installation (tank against failures that can lead to loss of life, material damage and further undesired costs of their unreliability. That is why a new, professional computer program has been created that allows for the selection of safety valves or for the verification of a safety valve selection in installations where any technical or technological changes have been made.

Transcatheter pulmonary valve replacement by hybrid approach using a novel polymeric prosthetic heart valve: proof of concept in sheep.

Directory of Open Access Journals (Sweden)

Ben Zhang

Full Text Available Since 2000, transcatheter pulmonary valve replacement has steadily advanced. However, the available prosthetic valves are restricted to bioprosthesis which have defects like poor durability. Polymeric heart valve is thought as a promising alternative to bioprosthesis. In this study, we introduced a novel polymeric transcatheter pulmonary valve and evaluated its feasibility and safety in sheep by a hybrid approach.We designed a novel polymeric trileaflet transcatheter pulmonary valve with a balloon-expandable stent, and the valve leaflets were made of 0.1-mm expanded polytetrafluoroethylene (ePTFE coated with phosphorylcholine. We chose glutaraldehyde-treated bovine pericardium valves as control. Pulmonary valve stents were implanted in situ by a hybrid transapical approach in 10 healthy sheep (8 for polymeric valve and 2 for bovine pericardium valve, weighing an average of 22.5±2.0 kg. Angiography and cardiac catheter examination were performed after implantation to assess immediate valvular functionality. After 4-week follow-up, angiography, echocardiography, computed tomography, and cardiac catheter examination were used to assess early valvular function. One randomly selected sheep with polymeric valve was euthanized and the explanted valved stent was analyzed macroscopically and microscopically.Implantation was successful in 9 sheep. Angiography at implantation showed all 9 prosthetic valves demonstrated orthotopic position and normal functionality. All 9 sheep survived at 4-week follow-up. Four-week follow-up revealed no evidence of valve stent dislocation or deformation and normal valvular and cardiac functionality. The cardiac catheter examination showed the peak-peak transvalvular pressure gradient of the polymeric valves was 11.9±5.0 mmHg, while that of two bovine pericardium valves were 11 and 17 mmHg. Gross morphology demonstrated good opening and closure characteristics. No thrombus or calcification was seen macroscopically

Dynamic testing of POSI-SEAL motor-operated butterfly valves using strain gages

International Nuclear Information System (INIS)

Richard, M.C.; Chiou, D.

1994-01-01

Utilities operating nuclear power plants recognize that the correct functioning of all motor-operated valves, and particularly those in safety-related systems, is of paramount importance. The U.S. Nuclear Regulatory Commission has issued Generic Letter 89-10 relative to this concern. Operability must be demonstrated under design-basis conditions. In order to demonstrate operability of motor-operated butterfly valves, the valve stem torque must be determined. The valve stem torque is a function of seat material, stem packing, stem bearing friction, and hydrodynamic lift and drag. The total valve operating hydrodynamic torque can be predicted using the valve manufacturer's data and the differential pressure. In order to validate the valve manufacturer's data, the actual total valve hydrodynamic torque is measured using strain gages mounted directly on the valve stem. This paper presents the results of comparing the predicted total valve operating hydrodynamic torque with the actual total valve operating hydrodynamic torque for six POSI-SEAL Class 150 high performance butterfly valves

Research on digital system design of nuclear power valve

Science.gov (United States)

Zhang, Xiaolong; Li, Yuan; Wang, Tao; Dai, Ye

2018-04-01

With the progress of China's nuclear power industry, nuclear power plant valve products is in a period of rapid development, high performance, low cost, short cycle of design requirements for nuclear power valve is proposed, so there is an urgent need for advanced digital design method and integrated design platform to provide technical support. Especially in the background of the nuclear power plant leakage in Japan, it is more practical to improve the design capability and product performance of the nuclear power valve. The finite element numerical analysis is a common and effective method for the development of nuclear power valves. Nuclear power valve has high safety, complexity of valve chamber and nonlinearity of seal joint surface. Therefore, it is urgent to establish accurate prediction models for earthquake prediction and seal failure to meet engineering accuracy and calculation conditions. In this paper, a general method of finite element modeling for nuclear power valve assembly and key components is presented, aiming at revealing the characteristics and rules of finite element modeling of nuclear power valves, and putting forward aprecision control strategy for finite element models for nuclear power valve characteristics analysis.

Quality and Safety in Health Care, Part XXX: Transcatheter Aortic Valve Therapy.

Science.gov (United States)

Harolds, Jay A

2017-12-01

Initially, the transcatheter aortic valve replacement procedure was approved only for patients with aortic stenosis that was both severe and symptomatic who either also had too high a risk of aortic valve replacement surgery to have the surgery or who had a high risk for the surgery. Between the years 2012 and 2015, the death rate at 30 days declined from an initial rate of 7.5% to 4.6%. There has also been more use of the transfemoral approach over the years. In 2016, the transcatheter aortic valve replacement was approved for patients with aortic stenosis at intermediate risk of surgery.

Directly acting spring loaded safety valves as shock reducing measure; Direkt wirkende, federbelastete Sicherheitsventile als Druckstossreduzierende Massnahme

Energy Technology Data Exchange (ETDEWEB)

Ismaier, A.; Schluecker, E. [Erlangen-Nuernberg Univ. (DE). Lehrstuhl fuer Prozessmaschinen und Anlagentechnik (IPAT)

2010-05-15

Hydraulic shocks as induced by fast closure of armatures or by sudden pump failures are massive impacts in piping systems and require extensive measures to absorb the generated load. Basically the avoidance of water hammers are preferable but in case of emergency shutdowns unavoidable hydraulic shocks have to be reduced by appropriate measures. The authors describe experiments with spring loaded safety valves as shock reducing measures. It was shown that the vale dimensions is essential for the efficacy. A realistic modeling is possible using the one-dimensional fluid mechanics code ROLAST.

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.)

29 CFR 1910.264 - Laundry machinery and operations.

Science.gov (United States)

2010-07-01

... provided to prevent injury or damage caused by fluid escaping from relief or safety valves if vented to the... or abrasion; and they shall be cautioned not to touch or eat food until their hands have been... instructions. (2) Mechanical—(i) Safety guards. (a) No safeguard, safety appliance, or device attached to, or...

Advancements in valve technology and industry lessons lead to improved plant reliability and cost savings

International Nuclear Information System (INIS)

Sharma, V.; Kalsi, M.S.

2005-01-01

Plant reliability and safety hinges on the proper functioning of several valves. Recent advancements in valve technology have resulted in new analytical and test methods for evaluating and improving valve and actuator reliability. This is especially significant in critical service applications in which the economic impact of a valve failure on production, outage schedules and consequential damages far surpasses the initial equipment purchase price. This paper presents an overview of recent advances in valve technology driven by reliability concerns and cost savings objectives without comprising safety in the Nuclear Power Industry. This overview is based on over 27 years of experience in supporting US and International nuclear power utilities, and contributing to EPRI, and NSSS Owners' Groups in developing generic models/methodologies to address industry wide issues; performing design basis reviews; and implementing plant-wide valve reliability improvement programs. Various analytical prediction software and hardware solutions and training seminars are now available to implement valve programs covering power plants' lifecycle from the construction phase through life extension and power up rate. These tools and methodologies can enhance valve-engineering activities including the selection, sizing, proper application, condition monitoring, failure analysis, and condition based maintenance optimization with a focus on potential bad actors. This paper offers two such examples, the Kalsi Valve and Actuator Program (KVAP) and Check Valve Analysis and Prioritization (CVAP) [1-3, 8, 9, 11-13]. The advanced, validated torque prediction models incorporated into KVAP software for AOVs and MOVs have improved reliability of margin predictions and enabled cost savings through elimination of unwarranted equipment modifications. CVAP models provides a basis to prioritize the population of valves recommended for preventive maintenance, inspection and/or modification, allowing

Performance of relief valves; Das Verhalten von Sicherheits-Drosselklappen

Energy Technology Data Exchange (ETDEWEB)

Schneider, K [Stuttgart Univ. (Germany). Inst. fuer Hydraulische Stroemungsmaschinen

1990-12-31

Safety armatures in hydropower systems have to ensure safe shut-off in case of pipeline ruptures. Excessive flow rates result in excessive loads on armature components. The contribution describes investigations carried out at Stuttgart University to determine the hydraulic conditions and loads on the safety armatures in case of pipeline rupture. (orig.) [Deutsch] Sicherheitsarmaturen in Wasserkraftanlagen haben die Aufgabe, im Falle eines Rohrbruches sicher zu schliessen. Dabei treten infolge der ueberhoehten Stroemungsgeschwindigkeiten aussergewoehnliche Beanspruchungen der Armaturenteile auf. In der vorliegenden Abhandlung wird gezeigt, wie die Untersuchungen der hydraulischen Verhaeltnisse im Rohrbruchfall in einer Wasserkraftanlage und die damit verbundene Ermittlung der massgeblichen Beanspruchungen der Sicherheitsarmaturen am Institut fuer Hydraulische Stroemungsmaschinen der Universitaet Stuttgart durchgefuehrt werden. (orig.)

A Main Steam Safety Valve (MSSV) With Fixed Blowdown According to ASME Section III,Part NC-7512

International Nuclear Information System (INIS)

Follmer, Bernhard; Schnettler, Armin

2002-01-01

In 1986, the NRC issued the Information Notice (IN) 86-05 'Main Steam Safety Valve test failures and ring setting adjustments'. Shortly after this IN was issued, the Code was revised to require that a full flow test has to be performed on each CL.2 MSSV by the manufacturer to verify that the valve was adjusted so that it would reach full lift and thus full relieving capacity and would re-close at a pressure as specified in the valve Design Specification. In response to the concern discussed in the IN, the Westinghouse Owners Group (WOG) performed extensive full flow testing on PWR MSSVs and found that each valve required a unique setting of a combination of two rings in order to achieve full lift at accumulation of 3% and re-closing at a blowdown of 5%. The Bopp and Reuther MSSV type SiZ 2507 has a 'fixed blowdown' i.e. without any adjusting rings to adjust the 'blowdown' so that the blowdown is 'fixed'. More than 1000 pieces of this type are successfully in nuclear power plants in operation. Many of them since about 25 years. Therefore it can be considered as a proven design. It is new that an optimization of this MSSV type SiZ 2507 fulfill the requirements of part NC-7512 of the ASME Section III although there are still no adjusting rings in the flow part. In 2000, for the Qinshan Candu unit 1 and 2 full flow tests were performed with 32 MSSV type SiZ 2507 size 8'' x 12'' at 51 bar saturated steam in only 6 days. In all tests the functional performance was very stable. It was demonstrated by recording the signals lift and system pressure that all valves had acceptable results to achieve full lift at accumulation of 3% and to re-close at blowdown of 5%. This is an advantage which gives a reduction in cost for flow tests and which gives more reliability after maintenance work during outage compared to the common MSSV design with an individual required setting of the combination of the two rings. The design of the type SiZ 2507 without any adjusting rings in the

MMS two-phase nonequilibrium pressurizer

International Nuclear Information System (INIS)

Oh, S.J.; Sursock, J.P.

1987-01-01

The pressurizer of a nuclear steam supply system establishes and maintains the nuclear plant primary loop pressure within the prescribed limit. It is a vertical cylindrical vessel which provides a water reserve and a steam surge chamber to accommodate coolant density changes during operation. To adjust the pressure to a desired value, electric heaters are provided in its lower section and the spray nozzles are provided in its upper section. Also, to protect against the buildup of the excess pressure, the pressurizer has two different types of relief valves, i.e., power operated relief valve and the safety relief valve. The pressurizer model implemented to the MMS is described in detail. In particular, the handling of the nonequilibrium condition, surgeline CCFL (Counter-current Flooding Limitation), and the level tracking model are described in detail. Next, the simulation of the Shippingport pressurizer load drop test is reported

Safety and efficacy of using the Viabahn endoprosthesis for percutaneous treatment of vascular access complications after transfemoral aortic valve implantation

DEFF Research Database (Denmark)

De Backer, Ole; Arnous, Samer; Sandholt, Benjamin

2015-01-01

Vascular access complications (VACs) remain one of the biggest challenges when performing transcatheter aortic valve implantation (TAVI). This study aimed to investigate the short- and medium-term safety and efficacy of the Viabahn endoprosthesis (Gore, Flagstaff, AZ) when used to treat TAVI......-induced vascular injury. Over a 40-month period, 354 patients underwent true percutaneous transfemoral (TF)-TAVI using a CoreValve and Prostar-XL closure system; this was our study population. A VAC leading to acute intervention occurred in 72 patients (20.3%) - of these, 18 were managed by balloon angioplasty, 48...... were treated by Viabahn stenting (technical success rate 98%), and 6 needed surgical intervention. Overall, this approach resulted in a major VAC rate of 3.1% (n = 11) in our study cohort. Length of hospitalization and 30-day mortality rates were comparable in patients with a VAC treated by Viabahn...

Safety analysis of high temperature reactor cooled and moderated by supercritical light water

International Nuclear Information System (INIS)

Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

2003-01-01

This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

Inspection systems for valves monitoring at EDF

International Nuclear Information System (INIS)

Germain, J.L.; Granal, L.; Provost, D.; Touillez, M.

1997-01-01

Electricite de France (EDF) makes increasing use of valve inspection systems to guarantee safety in its pressurized water reactor plants, improve plant availability and facilitate condition-based maintenance. A portable system known as SAMIR has been developed for inspection of motor-operated valves, and is now used on EDF's 900-MW sites. For its 1300-MW units, EDF has chosen a more complete system which enables measuring thrust on the valve stem during a maneuver, using a sensor mounted on the yoke. To detect internal vale leaks, an on-site assessment has demonstrated the economic benefits of acoustic emission techniques. EDF has equipped its sites with analog leak detection systems which may soon be replaced by a digital model now being developed. (authors)

Proactive pressure relief system management of life cycle and ageing in nuclear power plants

International Nuclear Information System (INIS)

Kolenc, J.; Ferrar, S.

2011-01-01

The last major power nuclear station built in North America was built when the Altair Company introduced the first microcomputer sparking the PC frenzy. It is safe to assume that there have been a great many changes since 1977 on both accounts. As the world's aging nuclear plants continue to be challenged with maintenance and replacement issues (obsolescence), as well making improvements within their facilities, proper pressure relief system management looms as a growing concern. This problem grows more acute as new engineering best practices are promulgated across industries and regulatory standards become more rigorous with much stricter enforcements. Unlike most pieces of operating equipment in a nuclear facility, pressure relief devices demand an extra level of consideration; as they form the 'last line of defense'. Combine the on-going obsolescence issue, with today's ever increasing demands for overall plant and public safety; pressure relief safety management will require increasing 'proactive' efforts to ensure safe facilities. This paper has been written to address some global pressure relief system management issues with respect the worlds aging nuclear facilities. This paper reflects findings we have discovered while conducting engineering pressure relief system audits on various nuclear power stations. It should be noted that these finding are not atypical of similar findings in pressure relief systems in the hydrocarbon processing world. (author)

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

Long-term performance of motor-operated valves

Energy Technology Data Exchange (ETDEWEB)

Scarbrough, T.G.

1996-12-01

The U.S. Nuclear Regulatory Commission (NRC) requires that motor-operated valves (MOVs) important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Despite these requirements, operating experience and research revealed problems with the performance of MOVs in operating nuclear power plants. In response to the concerns about MOV performance, the NRC issued Generic Letter (GL) 89-10, {open_quotes}Safety-Related Motor-Operated Valve Testing and Surveillance,{close_quotes} and its supplements. Most licensees have completed the aspects of their GL 89-10 programs associated with the review of MOV design bases, verification of MOV switch settings initially, testing of MOVs under design-basis conditions where practicable, and improvement of evaluations of MOV failures and necessary corrective action. Licensees are establishing processes to ensure that the long-term aspects of their MOV programs, such as periodic verification of MOV capability and the trending of MOV problems, are maintained. The NRC staff is developing a generic letter to address periodic verification of MOV design-basis capability.

Long-term performance of motor-operated valves

International Nuclear Information System (INIS)

Scarbrough, T.G.

1996-01-01

The U.S. Nuclear Regulatory Commission (NRC) requires that motor-operated valves (MOVs) important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Despite these requirements, operating experience and research revealed problems with the performance of MOVs in operating nuclear power plants. In response to the concerns about MOV performance, the NRC issued Generic Letter (GL) 89-10, open-quotes Safety-Related Motor-Operated Valve Testing and Surveillance,close quotes and its supplements. Most licensees have completed the aspects of their GL 89-10 programs associated with the review of MOV design bases, verification of MOV switch settings initially, testing of MOVs under design-basis conditions where practicable, and improvement of evaluations of MOV failures and necessary corrective action. Licensees are establishing processes to ensure that the long-term aspects of their MOV programs, such as periodic verification of MOV capability and the trending of MOV problems, are maintained. The NRC staff is developing a generic letter to address periodic verification of MOV design-basis capability

Description of leakage monitoring system at Angra 2 nuclear power plant primary circuit

International Nuclear Information System (INIS)

Costa, Lilian Rose Sobral da; Mendes, Jorge Eduardo de Souza

1999-01-01

This paper describes the Leakage Monitoring System installed in Angra 2 NPP. This system has the task of detecting, localizing and quantifying leaks in systems for which rupture preclusion is cited. These systems include the reactor coolant pressure boundary, the main steam and feedwater lines within the containment, and the main steam safety and relief valve station in the valve annex. (author)

The Development of the Safety Related Valve Class 1E Electrical Motor, the Target and the Results

International Nuclear Information System (INIS)

Saban, I.; Grgic, D.; Fancev, T.; Flegar, Lj.; Novosel, N.

1996-01-01

The development of the safety related valves class 1E electric motor is described. The design implemented in order to satisfy the 1E requirements, and a way in which related 1E standards are addressed, are shown. The development was realized in three stages. In the first stage eight motorettes were made and the insulation system was tested. In the second stage the motor was produced in accordance with producer's prototype QA program. In the third stage part of the testing of the produced motor was made. The results of the testing, finished until now, show that produced motor, as well as similarly produced motors, is able to perform its safety function in the design bases accident conditions as requested by class 1E requirements. The rest of the testing (LOCA test) can be made on the same or similar motor in the future. (author)

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.

Posterior urethral valves: 10 years audit of epidemiologic, diagnostic and therapeutic aspects in Yaoundé gynaeco-obstetric and paediatric hospital.

Science.gov (United States)

Tambo, Faustin Felicien Mouafo; Tolefac, Paul Nkemtendong; Ngowe, Marcelin Ngowe; Minkande, Jacqueline Ze; Mbouche, Landry; Guemkam, Georgette; Telelen, Neville Alemnju; Angwafo, Fru Fobuzshi; Sosso, Aurelien Maurice

2018-05-21

The incidence of posterior urethral valve (PUV) is estimated at 1:5000-1:8000 males. It is the most common paediatric urologic urgency and the most common cause of male obstructive uropathy and chronic renal failure in children. The study aimed to describe the experience of Yaoundé gynaeco-obstetrics and paediatric hospital in the management of PUV. Retrospectively, medical records were retrieved over a ten year period and all data recorded and analyzed for study objectives. Patients were called and evaluated for outcomes regarding morbidity and mortality. A total of 18 patients all males were managed over the ten year period, given prevalence of 13 cases/100,000 admissions and an admission rate of 2 per annum. The median age at presentation was 22 months and 13 (72.2%) participants presented late. Voiding urethrocystogram was done in all the participants where it showed dilated and elongated posterior urethral valves in 16 (88.9%) of the cases. Endoscopic valve ablation resulted in the relief of obstruction in all but 3 (16.7%) participants that had residual valves and 2 (11.2%) participants that had urethral stenosis. Type I valves were most common in 14 (78.0%) participants. The mean duration of follow up was 34.56 ± 21.47 months. Complications at final follow up were: 10 (55.6%) chronic renal failure, 2 (11.2%) end-stage renal failure. The case fatality rate was 5.6%. Many patients present late in our setting with already established complications. There is the need to counsel parents/guardians on the importance of long-term follow up after relief of obstruction.

Aortic valve-sparing surgery in Marfan syndrome.

Science.gov (United States)

Nachum, Eyal; Shinfeld, Amichay; Kogan, Alexander; Preisman, Sergey; Levin, Shany; Raanani, Ehud

2013-08-01

Patients with Marfan syndrome are referred for cardiac surgery due to root aneurysm with or without aortic valve regurgitation. Because these patients are young and frequently present with normal-appearing aortic cusps, valve sparing is often recommended. However, due to the genetic nature of the disease, the durability of such surgery remains uncertain. Between February 2004 and June 2012, 100 patients in our department suffering from aortic aneurysm with aortic valve regurgitation underwent elective aortic valve-sparing surgery. Of them, 30 had Marfan syndrome, were significantly younger (30 +/- 13 vs. 53 +/- 16 years), and had a higher percentage of root aneurysm, compared with ascending aorta aneurysm in their non-Marfan counterparts. We evaluated the safety, durability, clinical and echocardiographic mid-term results of these patients. While no early deaths were reported in either group, there were a few major early complications in both groups. At follow-up (reaching 8 years with a mean of 34 +/- 26 months) there were no late deaths, and few major late complications in the Marfan group. Altogether, 96% and 78% of the patients were in New York Heart Association functional class I-II in the Marfan and non-Marfan groups respectively. None of the Marfan patients needed reoperation on the aortic valve. Freedom from recurrent aortic valve regurgitation > 3+ was 94% in the Marfan patients. Aortic valve-sparing surgery in Marfan symdrome patients is safe and yields good mid-term clinical outcomes.

Cobalt reduction of NSSS valve hardfacings for ALARA

Energy Technology Data Exchange (ETDEWEB)

Kim, Joo Hak; Lee, Sang Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-07-01

This report informs NSSS designer that replacement of materials is one of the major means of ALARA implementation, and describes that NSSS valves with high-cobalt hardfacing are significant contributors to post-shutdown radiation fields caused by activation of cobalt-59 to cobalt-60. Generic procedures for implementing cobalt reduction programs for valves are presented. Discussions are presented of the general and specific design requirements for valve hardfacing in nuclear service. The nuclear safety issues involved with changing valve hardfacing materials are discussed. The common methods used to deposit hardfacing materials are described together with an explanation of the wear measurements. Wear resistance, corrosion resistance, friction coefficient, and mechanical properties of candidate hardfacing alloys are given. World-wide nuclear utility experience with cobalt-free hardfacing alloys is described. The use of low-cobalt or cobalt-free alloys in other nuclear plant components is described. 17 figs., 38 tabs., 18 refs. (Author).

Cobalt reduction of NSSS valve hardfacings for ALARA

International Nuclear Information System (INIS)

Kim, Joo Hak; Lee, Sang Sub

1994-07-01

This report informs NSSS designer that replacement of materials is one of the major means of ALARA implementation, and describes that NSSS valves with high-cobalt hardfacing are significant contributors to post-shutdown radiation fields caused by activation of cobalt-59 to cobalt-60. Generic procedures for implementing cobalt reduction programs for valves are presented. Discussions are presented of the general and specific design requirements for valve hardfacing in nuclear service. The nuclear safety issues involved with changing valve hardfacing materials are discussed. The common methods used to deposit hardfacing materials are described together with an explanation of the wear measurements. Wear resistance, corrosion resistance, friction coefficient, and mechanical properties of candidate hardfacing alloys are given. World-wide nuclear utility experience with cobalt-free hardfacing alloys is described. The use of low-cobalt or cobalt-free alloys in other nuclear plant components is described. 17 figs., 38 tabs., 18 refs. (Author)

MSSV Modeling for Wolsong-1 Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Moon, Bok Ja; Choi, Chul Jin; Kim, Seoung Rae [KEPCO EandC, Daejeon (Korea, Republic of)

2010-10-15

The main steam safety valves (MSSVs) are installed on the main steam line to prevent the overpressurization of the system. MSSVs are held in closed position by spring force and the valves pop open by internal force when the main steam pressure increases to open set pressure. If the overpressure condition is relieved, the valves begin to close. For the safety analysis of anticipated accident condition, the safety systems are modeled conservatively to simulate the accident condition more severe. MSSVs are also modeled conservatively for the analysis of over-pressurization accidents. In this paper, the pressure transient is analyzed at over-pressurization condition to evaluate the conservatism for MSSV models

49 CFR 195.428 - Overpressure safety devices and overfill protection systems.

Science.gov (United States)

2010-10-01

... reliability of operation for the service in which it is used. (b) In the case of relief valves on pressure... the operator notes in the manual required by § 195.402 why compliance with that part is not necessary...

46 CFR 38.10-15 - Safety relief valves-TB/ALL.

Science.gov (United States)

2010-10-01

.... plus the maximum flow rate of the cargo filling pipes or, (2) The vapors generated under fire exposure..., constructed and flow tested for capacity in conformance with subpart 162.017 or 162.018 of subchapter Q... excessive external pressure. (f) Void spaces between the primary and secondary barriers of nonpressure...

The Klinger hot gas double axial valve

International Nuclear Information System (INIS)

Kruschik, J.; Hiltgen, H.

1984-01-01

The Klinger hot gas valve is a medium controlled double axial valve with advanced design features and safety function. It was first proposed by Klinger early in 1976 for the PNP-Project as a containment shut-off for hot helium (918 deg. C and 42 bar), because a market research has shown that such a valve is not state of present techniques. In the first stage of development a feasibility study had to be made by detailed design, calculation and by basic experiments for key components in close collaboration with Interatom/GHT. This was the basis for further design, calculation, construction and experimental work for such a valve prototype within the new development contract. The stage of knowledge to that time revealed the following key priority development areas: Finite element stress analysis for the highly stressed high temperature main components; development of an insulation layout; Detailed experimental tests of functionally important structural components or units of the valve, partly at Klingers (gasstatic bearings, flexible metallic sealing element, aerodynamic and thermohydraulic tests), partly at Interatom (actuator unit and also gasstatic bearings), partly at HRB in Juelich (flexible metallic sealing system, aerodynamic and thermohydraulic tests); Design of a test valve for experimental work in the KVK (test circuit at Interatom) for evaluation of temperature distribution and reliability of operation; Design of a prototype and extensive testing in the KVK

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.

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

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

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.

Recent NRC research activities addressing valve and pump issues

Energy Technology Data Exchange (ETDEWEB)

Morrison, D.L.

1996-12-01

The mission of the U.S. Nuclear Regulatory Commission (NRC) is to ensure the safe design, construction, and operation of commercial nuclear power plants and other facilities in the U.S.A. One of the main roles that the Office of Nuclear Regulatory Research (RES) plays in achieving the NRC mission is to plan, recommend, and implement research programs that address safety and technical issues deemed important by the NRC. The results of the research activities provide the bases for developing NRC positions or decisions on these issues. Also, RES performs confirmatory research for developing the basis to evaluate industry responses and positions on various regulatory requirements. This presentation summarizes some recent RES supported research activities that have addressed safety and technical issues related to valves and pumps. These activities include the efforts on determining valve and motor-operator responses under dynamic loads and pressure locking events, evaluation of monitoring equipment, and methods for detecting and trending aging of check valves and pumps. The role that RES is expected to play in future years to fulfill the NRC mission is also discussed.

BWR control rod drive scram pilot valve monitoring system

International Nuclear Information System (INIS)

Soden, R.A.; Kelly, V.

1984-01-01

The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechancial works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the ''insert'' side of the control rod piston and vents the ''withdraw'' side of the piston causing the rods to insert during a scam. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a ''half scram'', a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

BWR control rod drive scram pilot valve monitoring program

International Nuclear Information System (INIS)

Soden, R.A.; Kelly, V.

1986-01-01

The control rod drive system in a Boiling Water Reactor is the most important safety system in the power plant. All components of the system can be verified except the solenoid operated, scram pilot valves without scramming a rod. The pilot valve mechanical works is the weak link to the control rod drive system. These pilot valves control the hydraulic system which applies pressure to the insert side of the control rod piston and vents the withdraw side of the piston causing the rods to insert during a scram. The only verification that the valve is operating properly is to scram the rod. The concern for this portion of the system is demonstrated by the high number of redundant components and complete periodic testing of the electrical circuits. The pilot valve can become hung-up through wear, fracture of internal components, mechanical binding, foreign material or chemicals left in the valve during maintenance, etc. If the valve becomes hung-up the electrical tests performed will not indicate this condition and scramming the rod is in jeopardy. Only an attempt to scram a rod will indicate the hung-up valve. While this condition exists the rod is considered inoperative. This paper describes a system developed at a nuclear power plant that monitors the pilot valves on the control rod drive system. This system utilizes pattern recognition to assure proper internal workings of the scram pilot valves to plant operators. The system is totally automatic such that each time the valve is operated on a half scram, a printout is available to the operator along with light indication that each of the 370 valves (on one unit of a BWR) is operating properly. With this monitoring system installed, all components of the control rod drive system including the solenoid pilot valves can be verified as operational without scramming any rods

Investigation of valve failure problems in LWR power plants

International Nuclear Information System (INIS)

1980-04-01

An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

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.)

Unmanned Aerial Vehicle Systems for Disaster Relief: Tornado Alley

Science.gov (United States)

DeBusk, Wesley M.

2009-01-01

Unmanned aerial vehicle systems are currently in limited use for public service missions worldwide. Development of civil unmanned technology in the United States currently lags behind military unmanned technology development in part because of unresolved regulatory and technological issues. Civil unmanned aerial vehicle systems have potential to augment disaster relief and emergency response efforts. Optimal design of aerial systems for such applications will lead to unmanned vehicles which provide maximum potentiality for relief and emergency response while accounting for public safety concerns and regulatory requirements. A case study is presented that demonstrates application of a civil unmanned system to a disaster relief mission with the intent on saving lives. The concept utilizes unmanned aircraft to obtain advanced warning and damage assessments for tornados and severe thunderstorms. Overview of a tornado watch mission architecture as well as commentary on risk, cost, need for, and design tradeoffs for unmanned aerial systems are provided.

Seismic assessment of the Pickering pressure relief duct

International Nuclear Information System (INIS)

Ghobarah, A.

1995-05-01

The objectives of the study are to examine the structural response of the Pickering pressure relief duct when subjected to earthquake ground motion and to estimate the seismic withstand capacity of various components of the structural system on the basis of performance criteria consistent with the safety function of the duct. (author). 24 refs., 16 tabs., 31 figs

Percutaneous mitral valve edge-to-edge repair

DEFF Research Database (Denmark)

Nickenig, Georg; Estevez-Loureiro, Rodrigo; Franzen, Olaf

2014-01-01

BACKGROUND: The use of transcatheter mitral valve repair (TMVR) has gained widespread acceptance in Europe, but data on immediate success, safety, and long-term echocardiographic follow-up in real-world patients are still limited. OBJECTIVES: The aim of this multinational registry is to present a...

Engineering and maintenance applied to safety-related valves in nuclear power plants

International Nuclear Information System (INIS)

Verdu, M. F.; Perez-Aranda, J.

2014-01-01

Nuclear Division in Iberdrola engineering and Construction has a team with extensive experience on engineering and services works related to valves. Also, this team is linked to UNESA as Technical support and Reference Center. Iberdrola engineering and construction experience in nuclear power plants valves, gives effective response to engineering and maintenance works that can be demanded in a nuclear power plant and it requires a high degree of qualification and knowledge both in Operation and Outages. (Author)

Operating experience of main steam isolation valves at Fessenheim and Bugey

International Nuclear Information System (INIS)

Dredemis, G.; Fourest, B.; Giroux, C.

1985-07-01

The paper presents the experience of Hopkinson MSIVs over about 40 reactor-years (1977 to 1984) of operation at Fessenheim and Bugey units (900 MWe PWR). The various problems encountered including ageing effects on auxiliary equipments and increases in closure time are discussed. The corrective actions undertaken by the utility and the safety assessment of these events performed by the french safety authorities are also described. This study is the synthesis of an in-depth analysis of Main Steam Isolation Valves (MSIV) and their auxiliary circuits equipping the Bugey and Fessenheim 900 MWe PWR nuclear power plants. These valves are different from those installed in the other French 900 MWe PWR reactors. The evaluation of the operation of these valves was made on the basis of incidents which occured during operation of the units or during the periodic tests, as well as anomalies discovered during maintenance operations. This analysis proved that the anomalies related to the design of the valves, as well as to their manufacture and installation, had been correctly dealt with. Furthermore, it should have also revealed potential anomalies due to ageing of the equipment

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)

Investigation of valve failure problems in LWR power plants

Energy Technology Data Exchange (ETDEWEB)

None

1980-04-01

An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

Removable butterfly valve, especially for polluting and/or dangerous fluids

International Nuclear Information System (INIS)

Garrigues, J.C.

1984-01-01

This invention relates to a valve for use in systems carrying polluting, corrosive or dangerous fluids requiring the use of biological protection. The facilities concerned are those in which fluids, mainly polluting, corrosive or dangerous liquids requiring the use of various types of biological protection, are handled. This is particularly so for nuclear installations in which the equipment is surrounded by protective shields which stop the radiation and prevent radioactive gases and aerosols from spreading. The invention proposes for the present valve a high-safety leaktightness system which respects the most specifications and standards and which suppresses any dead volume inside the valve [fr

Pressure wave propagation in the discharge piping with water pool

International Nuclear Information System (INIS)

Bang, Young S.; Seul, Kwang W.; Kim, In Goo

2004-01-01

Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined

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 ...

NRC perspective and experience on valve testing

International Nuclear Information System (INIS)

Eapen, P.K.

1990-01-01

Testing of safety related valves is one of the major activities at commercial nuclear power plants. In addition to Technical Specification, valve testing is required in 10 CFR 50.55a and 10 CFR 50 Appendix J. NRC inspectors (both resident and specialists) spend a considerable amount of time in following the valve test activities as part of their routine business. In the past, depending on a licensee's organizational structure, a valve could be tested more than three times to verify conformance with Technical Specifications, 10 CFR 50.55a, and 10 CFR 50 Appendix J. The regulatory reviewers were isolated from each other. Licensee test personnel were also not communicating among themselves. As a result, NRC inspectors found that certain valves in the IST program were inadequately tested. The typical licensee response was to say that this valve is exempted from testing under Appendix J. Others would say that the technical specification does not require fast closure of a valve in question. In addition to the above, the inspectors had to deal with exemption requests that were not dispositioned by the NRC. In the seventies there was a gentlemen's agreement to allow the licensee to do the testing in accordance with the exception, without waiting for the NRC approval. Needless to say when the new NRC inspection procedure was issued in March 1989 for implementation, the Regional inspectors had extremely difficult time to cope with the gray areas of valve testing. In August 1987, NRC Region I was reorganized and the special test program section was established to perform inspections in the IST area. This section was chartered to optimize resources and develop a meaningful inspection plan. The perspectives and insights used in the development of a detailed inspection plan is discussed below

Self-Expanding Transcatheter Aortic Valve System for Symptomatic High-Risk Patients With Severe Aortic Stenosis

DEFF Research Database (Denmark)

Reichenspurner, Hermann; Schaefer, Andreas; Schäfer, Ulrich

2017-01-01

BACKGROUND: The CENTERA transcatheter heart valve (THV) is a low-profile, self-expanding nitinol valve made from bovine pericardial tissue that is 14-F compatible with a motorized delivery system allowing for repositionability. OBJECTIVES: The pivotal study evaluated safety and efficacy of this THV...... permanent pacemaker was implanted in 4.5% of patients receiving the THV (4.9% for patients at risk). CONCLUSIONS: The herein described THV is safe and effective at 30 days with low mortality, significant improvements in hemodynamic outcomes, and low incidence of adverse events. Of particular interest...... is the low incidence of permanent pacemaker implantations. (Safety and Performance Study of the Edwards CENTERA-EU Self-Expanding Transcatheter Heart Valve [CENTERA-2]; NCT02458560)....

Adequacy of Flood Relief Shelters: A Case Study in Perak, Malaysia

Science.gov (United States)

Zawani Zahari, Nur; Mustafa Hashim, Ahmad

2018-03-01

The recent flood event occurred in 2014 had caused disastrous effects in Peninsular Malaysia in states of Kelantan, Pahang, Terengganu, Perak, Johor and Perlis. Perak state was reported with 12,115 victims from 2,896 families registered at 77 relief shelters. There are several issues encountered by the victims and related agencies which caused inconveniences and interruptions during the flooding period. Besides, the usage of public buildings as relief shelters contributes to deterioration of the infrastructures whereby their suitability, convenient, capacity and safety might not be optimum for longer period of time. This paper focuses on the assessment of relief shelters established in Perak Tengah district, Perak. Standards and guidelines for relief shelters were reviewed according to the most relevant agreed principles for humanitarian response. Data and information in this study were obtained from survey activities, interview sessions and observations. In Perak Tengah, more than 50% of the previous relief shelters were public buildings with low capacity areas. Strategic location of shelters with proper design standards should be established to ensure safe and healthy environment for the victims. Findings from this paper provide important outcomes to serve as better preparation in handling future disaster.

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

Safety of minimally invasive mitral valve surgery without aortic cross-clamp.

Science.gov (United States)

Umakanthan, Ramanan; Leacche, Marzia; Petracek, Michael R; Kumar, Sathappan; Solenkova, Nataliya V; Kaiser, Clayton A; Greelish, James P; Balaguer, Jorge M; Ahmad, Rashid M; Ball, Stephen K; Hoff, Steven J; Absi, Tarek S; Kim, Betty S; Byrne, John G

2008-05-01

We developed a technique for open heart surgery through a small (5 cm) right-anterolateral thoracotomy without aortic cross-clamp. One hundred and ninety-five consecutive patients (103 male and 92 female), age 69 +/- 8 years, underwent surgery between January 2006 and July 2007. Mean preoperative New York Heart Association function class was 2.2 +/- 0.7. Thirty-five patients (18%) had an ejection fraction 0.35 or less. Cardiopulmonary bypass was instituted through femoral (176 of 195, 90%), axillary (18 of 195, 9%), or direct aortic (1 of 195, 0.5%) cannulation. Under cold fibrillatory arrest (mean temperature 28.2 degrees C) without aortic cross-clamp, mitral valve repair (72 of 195, 37%), mitral valve replacement (117 of 195, 60%), or other (6 of 195, 3%) procedures were performed. Concomitant procedures included maze (45 of 195, 23%), patent foramen ovale closure (42 of 195, 22%) and tricuspid valve repair (16 of 195, 8%), or replacement (4 of 195, 2%). Thirty-day mortality was 3% (6 of 195). Duration of fibrillatory arrest, cardiopulmonary bypass, and "skin to skin" surgery were 88 +/- 32, 118 +/- 52, and 280 +/- 78 minutes, respectively. Ten patients (5%) underwent reexploration for bleeding and 44% did not receive any blood transfusions. Six patients (3%) sustained a postoperative stroke, eight (4%) developed low cardiac output syndrome, and two (1%) developed renal failure requiring hemodialysis. Mean length of hospital stay was 7 +/- 4.8 days. This simplified technique of minimally invasive open heart surgery is safe and easily reproducible. Fibrillatory arrest without aortic cross-clamping, with coronary perfusion against an intact aortic valve, does not increase the risk of stroke or low cardiac output. It may be particularly useful in higher risk patients in whom sternotomy with aortic clamping is less desirable.

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.

Brunswick Steam Electric Plant, Unit 2. Annual operating report, 1975

International Nuclear Information System (INIS)

1976-01-01

Initial criticality occurred in March and start-up testing continued for most of the year. Information is presented concerning operations, shutdowns, maintenance, design changes, personnel radiation exposures, thermal transients, safety/relief valve malfunctions, condenser tube leaks, and recirculation pump seal failures

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

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.

MOVES: A knowledge-based system for maintenance planning for motor-operated valves

International Nuclear Information System (INIS)

Winter, M.

1987-01-01

Over the past several years, knowledge-based expert systems have emerged as an important part of the general research area known as artificial intelligence. This paper describes a cooperative effort between faculty members at Iowa State University and engineers at the Duane Arnold Energy Center [a 545-MW(electric) boiling water reactor operated by Iowa Electric Light and Power Company] to explore the development of an advisory system for valve maintenance planning. This knowledge-based program, known as Motor-Operated Valves Expert System (MOVES), has a data base that currently includes safety-related motor-operated valves (∼117 valves). Valve maintenance was selected as the subject for the expert system because of the importance of valves in nuclear plant and their impact of plant availability. MOVES is being developed using the microcomputer-(IBM compatible) based expert system tool INSIGHT2+. The authors have found that the project benefits both the university and the utility

Residual stress measurement inside a dissimilar metal weld mock-up of the pressurizer safety and relief nozzle

International Nuclear Information System (INIS)

Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R.; Martins, Ketsia S.

2015-01-01

Residual stresses are present in materials or structural component in the absence of external loads or changes in temperatures. The most common causes of residual stresses being present are the manufacturing or assembling processes. All manufacturing processes, such as casting, welding, machining, molding, heat treatment, among others, introduces residual stresses into the manufactured object. The residual stresses effects could be beneficial or detrimental, depending on its distribution related to the component or structure, its load service and if it is compressive or tensile. In this work, the residual strains and stresses inside a mock-up that simulates the safety and relief nozzle of Angra 1 Nuclear Power Plant pressurizer were studied. The current paper presents a blind hole-drilling method residual stress measurements both at the inner surface of dissimilar metal welds of dissimilar metal weld nozzle mock-up. (author)

Residual stress measurement inside a dissimilar metal weld mock-up of the pressurizer safety and relief nozzle

Energy Technology Data Exchange (ETDEWEB)

Campos, Wagner R.C.; Rabello, Emerson G.; Silva, Luiz L.; Mansur, Tanius R., E-mail: wrcc@cdtn.br, E-mail: egr@cdtn.br, E-mail: silvall@cdtn.br, E-mail: tanius@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte (Brazil). Servico de Integridade Estrutural; Martins, Ketsia S., E-mail: ketshinoda@hotmail.com [Universidade Federal de Minas Gerais (UFMG), Nelo Horizonte (Brazil). Departamento de Engenharia Metalurgica

2015-07-01

Residual stresses are present in materials or structural component in the absence of external loads or changes in temperatures. The most common causes of residual stresses being present are the manufacturing or assembling processes. All manufacturing processes, such as casting, welding, machining, molding, heat treatment, among others, introduces residual stresses into the manufactured object. The residual stresses effects could be beneficial or detrimental, depending on its distribution related to the component or structure, its load service and if it is compressive or tensile. In this work, the residual strains and stresses inside a mock-up that simulates the safety and relief nozzle of Angra 1 Nuclear Power Plant pressurizer were studied. The current paper presents a blind hole-drilling method residual stress measurements both at the inner surface of dissimilar metal welds of dissimilar metal weld nozzle mock-up. (author)

Workshop on gate valve pressure locking and thermal binding

International Nuclear Information System (INIS)

Brown, E.J.

1995-07-01

The purpose of the Workshop on Gate Valve Pressure Locking and Thermal Binding was to discuss pressure locking and thermal binding issues that could lead to inoperable gate valves in both boiling water and pressurized water reactors. The goal was to foster exchange of information to develop the technical bases to understand the phenomena, identify the components that are susceptible, discuss actual events, discuss the safety significance, and illustrate known corrective actions that can prevent or limit the occurrence of pressure locking or thermal binding. The presentations were structured to cover U.S. Nuclear Regulatory Commission staff evaluation of operating experience and planned regulatory activity; industry discussions of specific events, including foreign experience, and efforts to determine causes and alleviate the affects; and valve vendor experience and recommended corrective action. The discussions indicated that identifying valves susceptible to pressure locking and thermal binding was a complex process involving knowledge of components, systems, and plant operations. The corrective action options are varied and straightforward

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

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

Probabilistic safety assessment of Narora Atomic Power Project

International Nuclear Information System (INIS)

Babar, A.K.; Saraf, R.K.; Kakodkar, A.; Sanyasi Rao, V.V.S.

1989-01-01

Various safety studies on Pressurised Water and Boiling Water reactors have been conducted. However, a detailed report on probabilistic safety assessment (PSA) of PHWRs is not available. PSA level I results of the standardised 235 MWe PHWR under construction at Narora are presented herein. Fault Tree analysis of various initiating events (IEs), safety systems has been completed. Event Tree analysis has been performed for all the dominating IEs to identify the accident sequences and a list of the dominating accident sequences is included. Analysis has been carried out using Monte Carlo simulation to propagate the uncertanities in failure rate data. Further uncertainty analysis is extended to obtain distributions for the accident sequences and core damage frequency. Some noteworthy results of the study apart from the various design modifications incorporated during the design phase are: (i) The accident sequences resulting from station blackout are dominant contributors to the core damage frequency. (ii) Class-IV transients, small break LOCA are significant IEs. Main steam line break is likely to induce steam generator tube ruptures. (iii) Moderator circulation, fire fighting system, secondary steam relief are relatively important in core damage frequency reductions. (iv) Under accidental situations human errors are likely to be asociated with valving in shutdown cooling and fire fighting systems. (author). 14 tabs., 14 figs., 15 refs

Components for containment enclosures. Part 4: Ventilation and gas-cleaning systems such as filters, traps, safety and regulation valves, control and protection devices

International Nuclear Information System (INIS)

2001-01-01

ISO 11933 consists of the following parts, under the general title Components for containment enclosures: Part 1: Glove/bag ports, bungs for glove/bag ports, enclosure rings and interchangeable units; Part 2: Gloves, welded bags, gaiters for remote-handling tongs and for manipulators; Part 3: Transfer systems such as plain doors, airlock chambers, double door transfer systems, leaktight connections for waste drums; Part 4: Ventilation and gas-cleaning systems such as filters, traps, safety and regulation valves, control and protection devices; Part 5: Penetrations for electrical and fluid circuits. This part of ISO 11933 specifies the design criteria and the characteristics of various components used for ventilation and gas-cleaning in containment enclosures. These components are either directly fixed to the containment enclosure wall, or used in the environment of a shielded or unshielded containment enclosure or line of such enclosures. They can be used alone or in conjunction with other mechanical components, including those specified in ISO 11933-1 and ISO 11933-3. This part of ISO 11933 is applicable to: filtering devices, including high-efficiency particulate air (HEPA) filters and iodine traps; safety valves and pressure regulators; systems ensuring the mechanical protection of containment enclosures; control and pressure-measurement devices

Alternative medicine - pain relief

Science.gov (United States)

Acupuncture - pain relief; Hypnosis - pain relief; Guided imagery - pain relief ... neck, shoulder, knee, or elbow) Osteoarthritis Rheumatoid arthritis Hypnosis is a focused state of concentration. With self- ...

Piezoelectric and Sensitivity Evaluation of Acoustic Emission Sensors for Nuclear Power Plant Valve

International Nuclear Information System (INIS)

Lee, Sang-Guk; Lee, Sun-Ki; Park, Sung-Keun; Kim, Myung-Ki

2008-01-01

A lot of valves are used in the power plant. The operation safety test and the valve inside leak detection are implemented on the valve which has a great impact on the safe operation of the plant. While input and output pressure measurement using a pressure gauge, temperature change and the humidity measurement, and pressure-resistant test are used for the valve leak detection, there are many problems such as the difficulty of the real time measurement at the minute leak situation, complexity of the pressure gauge correction and the process of the pressure measurement, and the reliability of the measured value. Therefore, it is necessary to develop the valve leak detection system using the acoustic emission (AE) method which is fast and accurate, and allows the real time measurement and evaluation of the minute leak situation. The valve leak detection method using the AE method is a convenient way to detect the sound of the leak outside the valve in case of existing leak inside of the valve, and the research is in progress recently to apply the method to the power plant valve

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 ...

Quad Cities Unit 2 Main Steam Line Acoustic Source Identification and Load Reduction

International Nuclear Information System (INIS)

DeBoo, Guy; Ramsden, Kevin; Gesior, Roman

2006-01-01

The Quad Cities Units 1 and 2 have a history of steam line vibration issues. The implementation of an Extended Power Up-rate resulted in significant increases in steam line vibration as well as acoustic loading of the steam dryers, which led to equipment failures and fatigue cracking of the dryers. This paper discusses the results of extensive data collection on the Quad Cities Unit 2 replacement dryer and the Main Steam Lines. This data was taken with the intent of identifying acoustic sources in the steam system. Review of the data confirmed that vortex shedding coupled column resonance in the relief and safety valve stub pipes were the principal sources of large magnitude acoustic loads in the main steam system. Modifications were developed in sub-scale testing to alter the acoustic properties of the valve standpipes and add acoustic damping to the system. The modifications developed and installed consisted of acoustic side branches that were attached to the Electromatic Relief Valve (ERV) and Main Steam Safety Valve (MSSV) attachment pipes. Subsequent post-modification testing was performed in plant to confirm the effectiveness of the modifications. The modifications have been demonstrated to reduce vibration loads at full Extended Power Up-rate (EPU) conditions to levels below those at Original Licensed Thermal Power (OLTP). (authors)

Expanding TAVI options: elective rotational atherectomy during trans-catheter aortic valve implantation

International Nuclear Information System (INIS)

Piccoli, Anna; Lunardi, Mattia; Ariotti, Sara; Ferrero, Valeria; Vassanelli, Corrado; Ribichini, Flavio

2015-01-01

Summary: Aortic valve stenosis (AVS) in the elderly is frequently associated to coronary artery disease (CAD). In patients with significant coronary stenosis surgical valve replacement is associated to coronary bypass grafting, but whether coronary angioplasty is needed in patients receiving trans-catheter aortic valve implantation (TAVI) is unknown. Given the frequent complexity of CAD in the elderly with calcific AVS, rotational atherectomy (RA) may be needed in some cases. No data are available about feasibility and safety of RA during TAVI. The need for myocardial revascularization in TAVI candidates is discussed, and a series of RA cases performed during TAVI is described

Expanding TAVI options: elective rotational atherectomy during trans-catheter aortic valve implantation

Energy Technology Data Exchange (ETDEWEB)

Piccoli, Anna; Lunardi, Mattia; Ariotti, Sara; Ferrero, Valeria; Vassanelli, Corrado; Ribichini, Flavio, E-mail: flavio.ribichini@univr.it

2015-01-15

Summary: Aortic valve stenosis (AVS) in the elderly is frequently associated to coronary artery disease (CAD). In patients with significant coronary stenosis surgical valve replacement is associated to coronary bypass grafting, but whether coronary angioplasty is needed in patients receiving trans-catheter aortic valve implantation (TAVI) is unknown. Given the frequent complexity of CAD in the elderly with calcific AVS, rotational atherectomy (RA) may be needed in some cases. No data are available about feasibility and safety of RA during TAVI. The need for myocardial revascularization in TAVI candidates is discussed, and a series of RA cases performed during TAVI is described.

Thermal-hydraulic analysis of total loss of steam generator feed water in WWER-440

International Nuclear Information System (INIS)

Sabotinov, L.; Cadet-Mercier, S.

2001-01-01

The analysis is carried out for a WWER-440/V270 with upgraded primary safety valves (replacement of the existing PRZ safety valves with Pilot Operated Relief Valves (PORV) of the type SEBIM (France)) The current analysis is focused on the scenario 'Total Loss of SGs Feed Water' with application of the operator action of primary system 'Feed and Bleed' in order to check the effectiveness of the installed pressurizer SEBIM valves and to verify that the operator can cool down the reactor system and cope with this accident. The calculations have been performed at the Institute of Protection and Nuclear Safety (IPSN) in Fontenay-aux-Roses with the computer code CATHARE 2 Version 1.3L1. CATHARE is a French best estimate thermal-hydraulic program for accident analysis in the light water nuclear reactors, developed with the participation of the IPSN (Institut de Protection et Surete Nucleaire), CEA (Commissariat a l'Energie Atomique), Framatome and EdF (Electricite de France). (author)

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

30 CFR 250.802 - Design, installation, and operation of surface production-safety systems.

Science.gov (United States)

2010-07-01

... Analysis Checklists are included in API RP 14C you must utilize the analysis technique and documentation... device requirements for pipelines are under § 250.1004. (c) Specification for surface safety valves (SSV..., Recommended Practice for Installation, Maintenance, and Repair of Surface Safety Valves and Underwater Safety...

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

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). ...

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...

Detection circuit of solenoid valve operation and control rod drive mechanism utilizing the circuit

International Nuclear Information System (INIS)

Ono, Takehiko.

1976-01-01

Object: To detect the operation of a plunger and detect opening and closing operations of a solenoid valve driving device due to change in impedance of a coil for driving the solenoid valve to judge normality and abnormality of the solenoid valve, thereby increasing reliance and safety of drive and control apparatus of control rods. Structure: An arrangement comprises a drive and operation detector section wherein the operation of a solenoid driving device for controlling power supply to a coil for driving the solenoid valve to control opening and closing of the solenoid valve, and a plunger operation detector section for detecting change in impedance of the drive coil to detect that the plunger of the solenoid valve is either in the opening direction or closing direction, whereby a predetermined low voltage such as not to activate the solenoid valve even when the solenoid valve is open or closed is applied to detect a current flowing into the coil at that time, thus detecting an operating state of the plunger. (Yoshino, Y.)

A method for evaluating pressure locking and thermal binding of gate valves

Energy Technology Data Exchange (ETDEWEB)

Dogan, T.

1996-12-01

A method is described to evaluate the susceptibility of gate valves to pressure locking and thermal binding. Binding of the valve disc in the closed position due to high pressure water trapped in the bonnet cavity (pressure locking) or differential thermal expansion of the disk in the seat (thermal binding) represents a potential mechanism that can prevent safety-related systems from functioning when called upon. The method described here provides a general equation that can be applied to a given gate valve design and set of operating conditions to determine the susceptibility of the valve to fail due to disc binding. The paper is organized into three parts. The first part discusses the physical mechanisms that cause disc binding. The second part describes the mathematical equations. The third part discusses the conclusions.

Multidetector computed tomography sizing of bioprosthetic valves: guidelines for measurement and implications for valve-in-valve therapies

International Nuclear Information System (INIS)

Rajani, R.; Attia, R.; Condemi, F.; Webb, J.; Woodburn, P.; Hodson, D.; Nair, A.; Preston, R.; Razavi, R.; Bapat, V.N.

2016-01-01

Aim: To describe a technique for bioprosthetic multidetector computed tomography (MDCT) sizing and to compare MDCT-derived values against manufacturer-provided sizing. Materials and methods: Fourteen bioprosthetic stented valves commonly used in the aortic valve position were evaluated using a Philips 256 MDCT system. All valves were scanned using a dedicated cardiac CT protocol with a four-channel electrocardiography (ECG) simulator. Measurements were made of major and minor axes and the area and perimeter of the internal stent using varying reconstruction kernels and window settings. Measurements derived from MDCT (MDCT ID) were compared against the stent internal diameter (Stent ID) as provided by the valve manufacturer and the True ID (Stent ID + insertion of leaflets). All data were collected and analysed using SPSS for Mac (version 21). Results: The mean difference between the MDCT ID and Stent ID was 0.6±1.9 mm (r=0.649, p=0.012) and between MDCT ID and True ID 2.1±2 mm (r=0.71, p=0.005). There was no difference in the major (p=0.90), minor (p=0.87), area (p=0.92), or perimeter (p=0.92) measurements when sharp, standard, and detailed stent kernels were used. Similarly, the measurements remained consistent across differing windowing levels. Conclusion: Bioprosthetic stented valves may be reliably sized using MDCT in patients requiring valve-in-valve (VIV) interventions where the valve type and size are unknown. In these cases, clinicians should be aware that MDCT has a tendency to overestimate the True ID size. - Highlights: • Cardiac CT is likely to be ideally suited for bioprosthetic aortic valve sizing for valve in valve procedures. • We compared MDCT sizing for 14 varying bioprosthetic aortic valves across varying window settings and reconstruction kernels. • We provide “normal” MDCT sizing for varying valves and show their relationship to surgical sizing. • Bioprosthetic valves may be reliably sized by MDCT but require adjustment owing to

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.)

RELAP5 Low Temperature Overpressurization Analysis for NPP Krsko

International Nuclear Information System (INIS)

Basic, I.; Krajnc, B.; Bajs, T.

2000-01-01

NEK has expressed interest in the acceptability of removing the autoclosure interlock (ACI) on the Residual Heat Removal System (RHRS) suction/isolation valves. This interest is in response to growing concerns about the loss of residual heat removal capability during cold shutdown and refueling operations. This article describes effort done to justify removal of the ACI in the light of low temperature overpressure protection of Reactor Coolant System (RCS) and RHRS and interfacing system LOCA potential. The intent of this article is to review the NEK RHRS relief valves sizing design basis and verify if the relief valves provide RHRS overpressure protection for the events possible at cold shutdown. Inadvertent isolation of RHRS during hot and cold shutdown (with reactor coolant system closed and temperatures below 177o Code 4 and5) presents one of the major safety concerns in this mode of operation. Detailed RELAP5 model of NPP Kriko following steam generator (SG) replacement and core uprate has been used in the frame of this analysis verification of RHRS relief valves sizing. The following limiting cases for cold shutdown with RCS solid conditions have been analyzed: - ransients that affect the system input/output mass balance, - ransients that affect the heat input/removal balance. (author)

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

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.

Venting of gas deflagrations through relief pipes

OpenAIRE

Ferrara, Gabriele

2006-01-01

Vent devices for gas and dust explosions are often ducted to safety locations by means of relief pipes for the discharge of hot combustion products or blast waves (NFPA 68, 2002). The presence of the duct is likely to increase the severity of the explosion with respect to simply vented vessels posing a problem for the proper design of this venting configuration. The phenomenology of the vented explosion is complicated as the interaction of combustion in the duct with primary combustion in...

A summary of the Chalk River valve packing evaluation program 1985 - 1990

International Nuclear Information System (INIS)

Aikin, J.A.; Doubt, G.L.; Lade, C.R.

1990-12-01

The move away from asbestos-based valve packing products has generated concern among valve manufacturers, packing manufacturers and user groups about the reliability and safety of non-asbestos based products for long-term use. AECL Research, Chalk River, has been actively evaluating these new valve packing products since 1985. This report describes the work done at Chalk River from 1985 to 1990. The report includes both Electric Power Research Institute (EPRI) and CANDU Owners Group (COG) funded studies. A description of the test programs and a brief summary of the functional performance of the more successful materials (die-formed graphite, braided asbestos and braided non-asbestos) on friction, stem leakage and consolidation are provided. At this time, Chalk River and Ontario Hydro have approved the following packing arrangements: for non-live-loaded valves, the recommended replacements packing for braided asbestos is combination flexible graphite packing sets; and, for heavy water valves originally designed with JC187I, the recommended replacement packing is approved braided-asbestos products

Acoustic emission condition monitoring of a nuclear power plant check valve using artificial neural networks

International Nuclear Information System (INIS)

Lee, Joon Hyun; Lee, Min Rae; Kim, Jung Teak

2005-01-01

In this study, an advanced condition monitoring technique based on acoustic emission (AE) detection and artificial neural networks was applied to a check valve, one of the components being used extensively in a safety system of a nuclear power plant (Npp). AE testing for a check valve under controlled flow loop conditions was performed to detect and evaluate disk movement for valve degradation such as wear and leakage due to foreign object interference in a check valve. It is clearly demonstrated that the evaluation of different types of failure modes such as disk wear and check valve leakage were successful by systematically analyzing the characteristics of various AE parameters. It is also shown that the leak size can be determined with an artificial neural network

Fatal association of mechanical valve thrombosis with dabigatran: a report of two cases.

Science.gov (United States)

Atar, Shaul; Wishniak, Alice; Shturman, Alexander; Shtiwi, Sewaed; Brezins, Marc

2013-07-01

Several new oral anticoagulants have been approved for thromboembolism prevention in patients with nonvalvular atrial fibrillation. However, they are not yet approved for anticoagulation use in patients with prosthetic mechanical valves, and no randomized data have been published so far on their safety of use in these patients. We present two cases of patients with prosthetic mechanical mitral valves who were switched from warfarin and acenocoumarol to dabigatran and within 1 month experienced severe valve complications resulting in death. One patient experienced stroke and later cardiogenic shock and death, and the other experienced pulmonary edema, cardiogenic shock, and subsequent death.

Results of safety analysis on PWR type nuclear power plants with two and three loops

International Nuclear Information System (INIS)

1979-01-01

The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

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’ ...

Nuclear power plant safety related pump issues

Energy Technology Data Exchange (ETDEWEB)

Colaccino, J.

1996-12-01

This paper summarizes of a number of pump issues raised since the Third NRC/ASME Symposium on Valve and Pump Testing in 1994. General issues discussed include revision of NRC Inspection Procedure 73756, issuance of NRC Information Notice 95-08 on ultrasonic flow meter uncertainties, relief requests for tests that are determined by the licensee to be impractical, and items in the ASME OM-1995 Code, Subsection ISTB, for pumps. The paper also discusses current pump vibration issues encountered in relief requests and plant inspections - which include smooth running pumps, absolute vibration limits, and vertical centrifugal pump vibration measurement requirements. Two pump scope issues involving boiling water reactor waterlog and reactor core isolation cooling pumps are also discussed. Where appropriate, NRC guidance is discussed.

Nuclear power plant safety related pump issues

International Nuclear Information System (INIS)

Colaccino, J.

1996-01-01

This paper summarizes of a number of pump issues raised since the Third NRC/ASME Symposium on Valve and Pump Testing in 1994. General issues discussed include revision of NRC Inspection Procedure 73756, issuance of NRC Information Notice 95-08 on ultrasonic flow meter uncertainties, relief requests for tests that are determined by the licensee to be impractical, and items in the ASME OM-1995 Code, Subsection ISTB, for pumps. The paper also discusses current pump vibration issues encountered in relief requests and plant inspections - which include smooth running pumps, absolute vibration limits, and vertical centrifugal pump vibration measurement requirements. Two pump scope issues involving boiling water reactor waterlog and reactor core isolation cooling pumps are also discussed. Where appropriate, NRC guidance is discussed

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.

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 ...

Dynamic performance of self-operated three-way valve used in a hybrid air conditioner

International Nuclear Information System (INIS)

Zhang, Penglei; Zhou, Dehai; Shi, Wenxing; Li, Xianting; Wang, Baolong

2014-01-01

A hybrid air conditioner combining a thermosyphon cycle with a vapor compression refrigeration cycle has a large energy saving potential compared with a common air conditioner for spaces requiring year-round cooling. The performance of the switch between the vapor compression mode and the thermosyphon mode largely impacts the safety and reliability of hybrid air conditioners. Therefore, a self-operated three-way valve is proposed. A thermodynamic model and a kinetic model are developed in this paper to evaluate the dynamic performance of the switch valve. The effects of the spring force constant, compressor discharging volume, fit clearance and piston length on the dynamic performance of the switch valve are analyzed. In conclusion, the proposed self-operated three-way valve can realize the switch operation accurately. - Highlights: •A self-operated three-way valve is proposed for hybrid air conditioners. •The thermodynamic model and kinetic model of the self-operated three-way valve are developed. •The validity of models is verified by experiments. •Effects of four main design parameters on the operating performance of the valve are researched

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

North American trial results at 1 year with the Sorin Freedom SOLO pericardial aortic valve.

Science.gov (United States)

Heimansohn, David; Roselli, Eric E; Thourani, Vinod H; Wang, Shaohua; Voisine, Pierre; Ye, Jian; Dabir, Reza; Moon, Michael

2016-02-01

A North American prospective, 15-centre Food and Drug Administration (FDA) valve trial was designed to assess the safety and effectiveness of the Freedom SOLO stentless pericardial aortic valve in the treatment of surgical aortic valve disease. Beginning in 2010, 251 patients (mean: 74.7 ± 7.5 years), were recruited in the Freedom SOLO aortic valve trial. One hundred eighty-nine patients have been followed for at least 1 year and are the basis for this review. Preoperatively, 54% of patients had NYHA functional class III or IV symptoms, and the majority of patients had a normal ejection fraction (EF) (median EF = 61%). Concomitant procedures were performed in 61.9% of patients, with coronary artery bypass grafting (CABG) (48.7%) being the most common followed by a MAZE procedure (13.7%). Reoperations were performed in 8.5% of patients in the study. The entire cohort of 251 patients enrolled had 7 deaths prior to 30 days, 2 of which were valve-related (aspiration pneumonia and sudden death) and 5 were not valve-related. There were 11 deaths after 30 days, 1 valve-related (unknown cardiac death) and 10 not valve-related. Five valves were explanted, 3 early (endocarditis, acute insufficiency and possible root dissection) and 2 late (endocarditis). Thirty-day adverse events include arrhythmias requiring permanent pacemaker (4.2%), thromboembolic events (3.7%) and thrombocytopenia (7.4%). One-year follow-up of all 189 patients demonstrated mean gradients for valve sizes 19, 21, 23, 25 and 27 mm of 11.7, 7.8, 6.3, 4.6 and 5.0 mmHg, respectively. Effective orifice areas for the same valve sizes were 1.2, 1.3, 1.6, 1.8 and 1.9 cm(2), respectively. Ninety-six percent of patients (181/189) were in NYHA class I or II at the 1-year follow-up. The Freedom SOLO stentless pericardial aortic valve demonstrated excellent haemodynamics and a good safety profile out to the 1 year of follow-up. © The Author 2015. Published by Oxford University Press on behalf of the European

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.)

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 ...

Structural valve deterioration in a starr-edwards mitral caged-disk valve prosthesis.

Science.gov (United States)

Aoyagi, Shigeaki; Tayama, Kei-Ichiro; Okazaki, Teiji; Shintani, Yusuke; Kono, Michitaka; Wada, Kumiko; Kosuga, Ken-Ichi; Mori, Ryusuke; Tanaka, Hiroyuki

2013-01-01

The durability of the Starr-Edwards (SE) mitral caged-disk valve, model 6520, is not clearly known, and structural valve deterioration in the SE disk valve is very rare. Replacement of the SE mitral disk valve was performed in 7 patients 23-40 years after implantation. Macroscopic examination of the removed disk valves showed no structural abnormalities in 3 patients, in whom the disk valves were removed at valves excised >36 years after implantation in 4 patients. Disk fracture, a longitudinal split in the disk along its circumference at the site of incorporation of the titanium ring, was detected in the valves removed 36 and 40 years after implantation, respectively, and many cracks were also observed on the outflow aspect of the disk removed 40 years after implantation. Disk fracture and localized disk wear were found in the SE mitral disk valves implanted >36 years previously. The present results suggest that SE mitral caged-disk valves implanted >20 years previously should be carefully followed up, and that those implanted >30 years previously should be electively replaced with modern prosthetic valves

Presentation, management, and outcome of posterior urethral valves in a Nigerian tertiary hospital.

Science.gov (United States)

Orumuah, Agbugui Jude; Oduagbon, Obarisiagbon Edwin

2015-01-01

Posterior urethral valves (PUV) remain the most common cause of bladder outlet obstruction and renal insufficiency in male children. The aim of this study was to evaluate the presentation, management, challenges, and outcome of the disease in a Nigerian tertiary health institution. Retrospectively, medical records of male children with a diagnosis of PUVs over a 10 year period (2003-2012) were retrieved. All data in relation to the study objectives were recorded and analyzed. A total of 44 patients was managed for PUV within the period. The mean age of presentation was 3.95 years with 56.8% of the patients presenting after the age of 1 year. Voiding dysfunction noted in 40 (91.0%) patients was the most common mode of presentation. The most common finding on physical examination was a palpable bladder while urinary tract infection noted in 23 (52.3%) patients was the most common complication noted. Abdominal ultrasonography revealed dilated posterior urethra in 16 (36.4%) cases, while micturating cystourethrogram revealed a dilated proximal urethra in all 35 cases in which it was done, diverticulum in 6 and vesicoureteric reflux in 9. The creatinine value at presentation ranged between 0.4 mg/dl and 4.0 mg/dl with a mean of 1.02 ± 0.93 mg/dl. Urethroscopy in 37 patients confirmed type I and type III PUV in 35 and 2 patients, respectively. Valve ablation with a diathermy bugbee electrode provided relief of obstructions in the 37 patients who underwent the procedure without any significant immediate complication. The period of follow-up ranged between 2 weeks and 3 years with a mean of 10.2 months. There was sustained improvement in urine stream, reduction in the mean creatinine concentration and incidence of UTI during follow-up. However, patients with significantly impaired renal function had a poorer outcome. Many patients with PUV presented late within the reviewed period. Valve ablation provided relief of obstruction in most of the cases. There is a need to

A Study of System Pressure Transients Generated by Isolation Valve Open/Closure in Orifice Manifold

Energy Technology Data Exchange (ETDEWEB)

Kim, M. [KEPCO, Daejeon (Korea, Republic of); Bae, S. W.; Kim, J. I.; Park, S. J. [KHNP, Abu Dhabi (United Arab Emirates)

2016-05-15

In this study, we explore the effects of pressure transients on peak and minimal pressures caused by the actuation of isolation valve and control valve reacting to the combined orifice operation of orifice manifold with motor-operated valve installed in the rear of the orifice. We then use the collected data to direct our effort towards cause analysis and propose improvements to efficiency and safety of operation. This formation is used to by domestic and foreign nuclear power plants as a mean to control flow rate, producing required flow rate jointly together by combination of the orifices. No significant impacts on the internals of manifold orifice due to peak pressure has been observed, although chance of cavitation at the outlet of control valve is significant. Considering the peak pressure, as well as minimum pressure occurs in low flow rate conditions, the pressure transient is more so affected by the characteristics (modified equal percentage) of control valve. Isolation valve of the orifice and control valve operate organically, therefore stroke time for valves need to be applied in order for both valves to cooperatively formulate an optimized operation.

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.

Dynamics of cold helium flow inside a cryoline used for large cryogenic distribution system

International Nuclear Information System (INIS)

Kumar, Uday; Jadon, Mohit; Choukekar, Ketan; Shukla, Vinit; Patel, Pratik; Kapoor, Himanshu; Shah, Nitin; Muralidhara, Srinivasa; Sarkar, Biswanath

2015-01-01

The Cryolines, which by definition transfers cryogens from the source, normally a cryogenic plant, to several systems requiring cooling at cryogenic temperature to the level of 4 K and 80 K. The operations of cryolines are normally assumed to be steady state following a cool down from room temperature to the cryogenic temperature. It is to be noted that in a distributed cryogenic system, especially in a nuclear facility such as ITER having confinement definition due to the regulatory requirements, do also attract the attention in the system design that the release from safety valves cannot be allowed inside a building. Therefore, all safety valves need to be discharged inside a confined space, which is a specific space requiring fulfillment of definition for a cryogenic line. The specificity in such cases is that such cryogenic lines will realize dynamic conditions for each release of safety valves or a combination of safety valves in terms of pressure, temperature and flow, leading to unexpected failures. Such operating scenarios also lead to serious impact on fatigue with a question mark on the reliability. Therefore, one can define such cryolines as Relief Collection Header (RCH) which collects discharged helium and transport it to the appropriate place as defined in the system design. The discharges of cold helium from safety relief discharge ports of equipment can result into significantly unsteady and compressible flow in RCH. The proper design of the RCH has to be supported by detailed dynamic of expected flow phenomena for specific cases. The paper presents the dynamics of cold helium flow inside the typical RCH that has been performed to investigate the variation in flow parameters (pressure, temperature, velocity and density) along the axis of RCH and predictions on its reliability. (author)

The JUPITER registry: 1-year results of transapical aortic valve implantation using a second-generation transcatheter heart valve in patients with aortic stenosis.

Science.gov (United States)

Silaschi, Miriam; Treede, Hendrik; Rastan, Ardawan J; Baumbach, Hardy; Beyersdorf, Friedhelm; Kappert, Utz; Eichinger, Walter; Rüter, Florian; de Kroon, Thomas L; Lange, Rüdiger; Ensminger, Stephan; Wendler, Olaf

2016-11-01

Transcatheter aortic valve replacement (TAVR) is an established therapy for patients with aortic stenosis (AS) at high surgical risk. The JenaValve™ is a second-generation, self-expanding transcatheter heart valve (THV), implanted through transapical access (TA). During stent deployment, a specific 'clipping-mechanism' engages native aortic valve cusps for fixation. We present 1-year outcomes of the JUPITER registry, a post-market registry of the JenaValve for TA-TAVR. The JUPITER registry is a prospective, multicentre, uncontrolled and observational European study to evaluate the long-term safety and effectiveness of the Conformité Européenne-marked JenaValve THV. A total of 180 patients with AS were enrolled between 2012 and 2014. End-points were adjudicated in accordance with the valve academic research consortium document no. 1 definitions. The mean age was 80.4 ± 5.9 years and the mean logistic European system for cardiac operative risk evaluation I 21.2 ± 14.7%. The procedure was successful in 95.0% (171/180), implantation of a second THV (valve-in-valve) was performed in 2.2% (4/180) and conversion to surgical aortic valve replacement (SAVR) was necessary in 2.8% (5/180). No annular rupture or coronary ostia obstruction occurred. Two patients required SAVR after the day of index procedure (1.1%). All-cause mortality at 30 days was 11.1% (20/180), being cardiovascular in 7.2% (13/180). A major stroke occurred in 1.1% (2/180) at 30 days, no additional major strokes were observed during 1 year. All-cause mortality after 30 days was 13.1% (21/160) and combined efficacy at 1 year was 80.8% (122/151). At 1-year follow-up, no patient presented with more than moderate paravalvular leakage, while 2 patients (3.2%) showed moderate, 12 (19.0%) mild and 49 (82.4%) trace/none paravalvular regurgitation. In a high-risk cohort of patients undergoing TA-TAVR for AS, the use of the JenaValve THV is safe and effective. In patients at higher risk for coronary ostia

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

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)

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.

The analysis of actuating mechanism and review of concepts for the vortex valve

International Nuclear Information System (INIS)

Park, Jong Kyun; Sim, Yun Seop; Joung, Sae Won; Lee, Ki Young; Lee, Jun; Kim, Young In

1995-12-01

To understand the basic features of the passive fluidic device, which is increasing available core cooling water from the safety injection tanks in the KNGR, review of the existing vortex valves concepts and analysis of the actuating mechanism of them have been performed and the results are as following: * Preliminary methodology development for parallel two water columns behavior, which is similar to the SIT valve actuation condition * Preliminary methodology for the vortex value actuation features * Analysis of the parallel water columns behavior and vortex valve actuation features using the results of above activities * Further works to be done in the analytical methodology. 16 figs., 2 refs. (Author) .new

Secure and effective valve stem sealing in PWR power generating plants

International Nuclear Information System (INIS)

Reynolds, J.

1991-01-01

The PWR power generating plant combines severe operating conditions with the highest safety requirements, making it one of the most demanding environments for seals. An analysis of the conditions inherent in its operation reveals: an aggressive and radioactive fluid at high temperature and pressure; frequent thermal shocks; and hazards for maintenance personnel in the containment area unless the reactor is shut down. The achievement of today's quality and safety standards owes much to the experience, research and testing carried out by the Electricite de France during its graduation from its first nuclear unit to become the world's most important manager of PWR plants with over 45 now under its control. The number of valves involved in the French nuclear program is in excess of 1,300,000. Knowing what the affect of a leak can be, especially if it necessitates a shutdown of the power station, the need to insure the quality of valve sealing can be appreciated. At the beginning of their nuclear building program, the EdF was finding that valves, representing only 2 percent of the investment in a PWR plant, caused 20% of the unwanted outages and cost 60% of the total of plant maintenance. In this report, the author endeavors to show how this problem was solved by team work and concerted action by the EdF, the valve constructors and seal manufacturer, not forgetting the importance of informing and training the maintenance and repair teams within the power stations themselves

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)

Nuclear power plant with a safety enclosure

International Nuclear Information System (INIS)

Keller, W.; Krueger, J.; Ropers, J.; Schabert, H.P.

1976-01-01

A nuclear power plant has a safety enclosure for a nuclear reactor. A fuel element storage basin is also located in this safety enclosure and a fuel element lock extends through the enclosure, with a cross-sectional size proportioned for the endwise passage of fuel elements, the lock including internal and external valves so that a fuel element may be locked endwise safely through the lock. The lock, including its valves, being of small size, does not materially affect the pressure resistance of the safety enclosure, and it is more easily operated than a lock large enough to pass people and fuel element transport vessels

Implantation and 30-Day Follow-Up on All 4 Valve Sizes Within the Portico Transcatheter Aortic Bioprosthetic Family

DEFF Research Database (Denmark)

Möllmann, Helge; Linke, Axel; Holzhey, David M

2017-01-01

OBJECTIVES: The aim of this study was to evaluate the short-term safety and performance of the full range of valve sizes offered within the Portico transcatheter aortic valve replacement system. BACKGROUND: The Portico transcatheter aortic heart valve is a fully resheathable, repositionable.......8% of patients improved by ≥1 New York Heart Association functional class at 30 days. The rate of moderate paravalvular leak was 5.7%, with no severe paravalvular leak reported. No differences in paravalvular leak incidence and severity were observed among valve sizes (p = 0.24). CONCLUSIONS: Across all valve...... sizes, use of the repositionable Portico transcatheter aortic valve replacement system resulted in safe and effective treatment of aortic stenosis in high-risk patients....

Role of recent research in improving check valve reliability at nuclear power plants

International Nuclear Information System (INIS)

Kalsi, M.S.; Horst, C.L.; Wang, J.K.; Sharma, V.

1990-01-01

Check valve failures at nuclear power plants in recent years have led to serious safety concerns, and caused extensive damage to other plant components which had a significant impact on plant availability. In order to understand the failure mechanism and improve the reliability of check valves, a systematic research effort was proposed by Kalsi Engineering, Inc. to U.S. Nuclear Regulatory Commission (NRC). The overall goal of the research was to develop models for predicting the performance and degradation of swing check valves in nuclear power plant systems so that appropriate preventive maintenance or design modifications can be performed to improve the reliability of check valves. Under Phase I of this research, a large matrix of tests was run with instrumented swing check valves to determine the stability of the disc under a variety of upstream flow disturbances, covering a wide range of disc stop positions and flow velocities in two different valve sizes. The goals of Phase II research were to develop predictive models which quantify the anticipated degradation of swing check valves that have flow disturbances closely upstream of the valve and are operating under flow velocities that do not result in full disc opening. This research allows the inspection/maintenance activities to be focussed on those check valves that are more likely to suffer premature degradation. The quantitative wear and fatigue prediction methodology can be used to develop a sound preventive maintenance program. The results of the research also show the improvements in check valve performance/reliability that can be achieved by certain modifications in the valve design

Design criteria of integrated reactors based on transients

International Nuclear Information System (INIS)

Zanocco, P.; Gimenez, M.; Delmastro, D.

1999-01-01

A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author)

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.

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

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.

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.

THE RESULTS OF SURGICAL TREATMENT OF TRICUSPID VALVE INFECTIVE ENDOCARDITIS USING VALVE REPAIR AND VALVE REPLACEMENT OPERATIONS

Directory of Open Access Journals (Sweden)

S. A. Kovalev

2015-01-01

Full Text Available Aim. To evaluate in-hospital and long-term results of surgical treatment of patients with infective endocarditis of the tricuspid valve, to compare the effectiveness of valve repair and valve replacement techniques, and to identify risk factors of mortality and reoperations. Materials and methods. 31 surgical patients with tricuspid valve infective endocarditis were evaluated. Patients were divided into 2 groups. In Group 1 (n = 14 repairs of the tricuspid valve were performed, in Group 2 (n = 17 patients had undergone tricuspid valve replacements. Epidemiological, clinical, microbiological and echocardiographic data were studied. Methods of comparative analysis, the Kaplan–Meier method, and Cox risk models were applied. Results. The most common complication of in-hospital stay was atrioventricular block (17.7% of cases in Group 2. In Group 1, this type of complication was not found. Hospital mortality was 7.14% in Group 1, and 0% in Group 2. Long-term results have shown the significant reduction of heart failure in general cohort and in both groups. In Group 1 the severity of heart failure in the long term was less than in Group 2. No significant differences in the severity of tricuspid regurgitation were found between the groups. In 7-year follow up no cases of death were registered in Group 1. Cumulative survival rate in Group 2 within 60 months was 67.3 ± 16.2%. No reoperations were performed in patients from Group 1. In Group 2, the freedom from reoperation within 60 months was 70.9 ± 15.3%. Combined intervention was found as predictor of postoperative mortality. Prosthetic valve endocarditis was identified as risk factor for reoperation. Conclusion. Valve repair and valve replacement techniques of surgical treatment of tricuspid valve endocarditis can provide satisfactory hospital and long-term results. Tricuspid valve repair techniques allowed reducing the incidence of postoperative atrioventricular block. In the long-term, patients

Clinical Implication of Transaortic Mitral Pannus Removal During Repeat Cardiac Surgery for Patients With Mechanical Mitral Valve.

Science.gov (United States)

Park, Byungjoon; Sung, Kiick; Park, Pyo Won

2018-01-25

This study aimed to evaluate the safety and feasibility of transaortic mitral pannus removal (TMPR).Methods and Results:Between 2004 and 2016, 34 patients (median age, 57 years; 30 women) with rheumatic disease underwent pannus removal on the ventricular side of a mechanical mitral valve through the aortic valve during reoperation. The median time interval from the previous surgery was 14 years. TMPR was performed after removal of the mechanical aortic valve (n=21) or diseased native aortic valve (n=11). TMPR was performed in 2 patients through a normal aortic valve. The mitral transprosthetic mean pressure gradient (TMPG) was ≥5 mmHg in 11 patients, including 3 with prosthetic valve malfunction. Prophylactic TMPR was performed in 23 patients. There were no early deaths. Concomitant operations included 22 tricuspid valve surgeries (13 replacements, 15 repairs) and 32 aortic valve replacements (24 repeats, 8 primary). The mean gradient in patients who had mitral TMPG ≥5 mmHg was significantly decreased from 6.46±1.1 to 4.37±1.17 mmHg at discharge (Ppannus overgrowth in such valves.

Chemical Safety Alert: Shaft Blow-Out Hazard of Check and Butterfly Valves

Science.gov (United States)

Certain types of check and butterfly valves can undergo shaft-disk separation and fail catastrophically, even when operated within their design limits of pressure and temperature, causing toxic/flammable gas releases, fires, and vapor cloud explosions.

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

NRC Information No. 88-72: Inadequacies in the design of dc motor-operated valves

International Nuclear Information System (INIS)

Rossi, C.E.

1992-01-01

On July 1, 1988, a high pressure coolant injection (HPCI) steam admission valve failed to open during a post-maintenance test at the Brunswick nuclear power plant, Unit 1. The same valve had failed in December 1987 and on May 28, 1988. The licensee, Carolina Power and Light Company, established a team to investigate the cause of failure, and the team identified the most probable cause as a dc motor failure due to a shunt-winding to series-winding short circuit. The team believed that this condition was precipitated by thermal binding of the valve internals. The previous failure in May was also diagnosed as having been caused by thermal binding. As a result of these failures, the licensee reviewed the design of the dc motor-operated valves for both the HPCI and the reactor core isolation cooling (RCIC) systems. This review identified a number of significant design deficiencies going well beyond the problems with thermal binding. The deficiencies constitute a potential common cause failure mechanism for safety system valves. Unit 1 was shut down on July 14, 1988 to replace the failed HPCI valve motor and to implement design modifications to other motor-operated valves

Blood Pressure and Arterial Load After Transcatheter Aortic Valve Replacement for Aortic Stenosis.

Science.gov (United States)

Lindman, Brian R; Otto, Catherine M; Douglas, Pamela S; Hahn, Rebecca T; Elmariah, Sammy; Weissman, Neil J; Stewart, William J; Ayele, Girma M; Zhang, Feifan; Zajarias, Alan; Maniar, Hersh S; Jilaihawi, Hasan; Blackstone, Eugene; Chinnakondepalli, Khaja M; Tuzcu, E Murat; Leon, Martin B; Pibarot, Philippe

2017-07-01

After aortic valve replacement, left ventricular afterload is often characterized by the residual valve obstruction. Our objective was to determine whether higher systemic arterial afterload-as reflected in blood pressure, pulsatile and resistive load-is associated with adverse clinical outcomes after transcatheter aortic valve replacement (TAVR). Total, pulsatile, and resistive arterial load were measured in 2141 patients with severe aortic stenosis treated with TAVR in the PARTNER I trial (Placement of Aortic Transcatheter Valve) who had systolic blood pressure (SBP) and an echocardiogram obtained 30 days after TAVR. The primary end point was 30-day to 1-year all-cause mortality. Lower SBP at 30 days after TAVR was associated with higher mortality (20.0% for SBP 100-129 mm Hg versus 12.0% for SBP 130-170 mm Hg; P <0.001). This association remained significant after adjustment, was consistent across subgroups, and confirmed in sensitivity analyses. In adjusted models that included SBP, higher total and pulsatile arterial load were associated with increased mortality ( P <0.001 for all), but resistive load was not. Patients with low 30-day SBP and high pulsatile load had a 3-fold higher mortality than those with high 30-day SBP and low pulsatile load (26.1% versus 8.1%; hazard ratio, 3.62; 95% confidence interval, 2.36-5.55). Even after relief of valve obstruction in patients with aortic stenosis, there is an independent association between post-TAVR blood pressure, systemic arterial load, and mortality. Blood pressure goals in patients with a history of aortic stenosis may need to be redefined. Increased pulsatile arterial load, rather than blood pressure, may be a target for adjunctive medical therapy to improve outcomes after TAVR. URL: http://www.clinicaltrials.gov. Unique identifier: NCT00530894. © 2017 American Heart Association, Inc.

Study of laser cladding nuclear valve parts

International Nuclear Information System (INIS)

Shi Shihong; Wang Xinlin; Huang Guodong

1998-12-01

The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO 2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

Double-walled control valves for the transport of liquids presenting a water pollution hazard; Doppelwandige Stellventile fuer den Transport wassergefaehrdender Fluessigkeiten

Energy Technology Data Exchange (ETDEWEB)

Daume, A.; Weissberg, S. [Daume Regelarmaturen GmbH, Isernhagen (Germany)

2004-09-01

Under German law valves, vessels and connecting pipework containing and/or transporting hazardous substances must be fitted with watertight drip pans or moniterable double walls. This article describes double-walled control valves which are very well suited to meet plant operators' safety requirements and environmental protection requirements. In addition to environmental protection, the valves provide opportunities for cost savings. (orig.)

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.

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.

Mechanical versus bioprosthetic mitral valve replacement in patients <65 years old.

Science.gov (United States)

Kaneko, Tsuyoshi; Aranki, Sary; Javed, Quratulain; McGurk, Siobhan; Shekar, Prem; Davidson, Michael; Cohn, Lawrence

2014-01-01

Because of its durability, the mechanical valve is typically chosen for young patients undergoing mitral valve replacement (MVR). However, a bioprosthetic valve might have the benefit of valve-in-valve transcatheter valve replacement when valve failure occurs. We examined the outcomes in patients who had undergone mechanical valve MVR (MVRm) versus bioprosthetic valve MVR (MVRb) in patients aged Security Death Index. The postoperative and long-term outcomes of interest included combined stroke and embolic events, reoperations, and mortality. Of 768 consecutive patients, 627 were in the MVRm and 141 in the MVRb group. Propensity score matching yielded a cohort of 125 MVRb (89%) and 125 control MVRm patients with similar etiology mixes. The groups were similar in age (MVRm, 53.2 ± 9.0 years; MVRb, 53.8 ± 10.6 years; P = .617) and other preoperative characteristics. The postoperative outcomes were also similar between the 2 groups, including reoperation for bleeding, stroke, deep sternal infection, sepsis, and length of hospital stay. The operative mortality was also similar (MVRm, 5.6%; MVRb, 8.0%; P = .617). However, Kaplan-Meier analysis showed the MVRb group had a greater reoperation rate (P = .001) and shorter estimated survival (11.3 vs 13.5 years, P = .004). The incidence of bleeding and stroke or embolic events between the 2 groups was similar. In the present report, MVRb for patients safety of mechanical valves in this group. Copyright © 2014 The American Association for Thoracic Surgery. Published by Mosby, Inc. All rights reserved.

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

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

Prevention of water hammers behind fast acting shut-off valves

International Nuclear Information System (INIS)

Dudlik, A.; Schlueter, S.; Prasser, H.M.

1999-01-01

Shut-off valves are often applied for a quick safety shut-down of pipelines for dangerous liquids. When the valve is suddenly closed at a certain flow rate, the pressure behind the valve decreases due to the momentum of the liquid. This leads to an evaporation and to a formation of a cavitation bubble if the resulting minimum pressure becomes lower than the saturation pressure at the given temperature. This type of water hammer was studied at a test facility of Fraunhofer UMSICHT in Oberhausen. The task of FZR was the development of an adequate two-phase flow instrumentation to visualise the cavitation phenomenon. During the project unique digital movie sequences of appearing and condensing cavitation bubbles were recorded, that gave valuable insights into the transient process. In the result, a new method was found to prevent the collapse of the cavitation bubbles and to avoid the corresponding pressure peaks. (orig.)

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.

Studies for the requirements of automatic and remotely controlled shutoff valves on hazardous liquids and natural gas pipelines with respect to public and environmental safety

Energy Technology Data Exchange (ETDEWEB)

Oland, C. Barry [XCEL Engineering, Inc. (United States); Rose, Simon D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Engineering Science and Technology Div.; Grant, Herb L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fabrication, Hoisting and Rigging Div.; Lower, Mark D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fabrication, Hoisting and Rigging Div.; Spann, Mark A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Facility Management Div.; Kirkpatrick, John R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Computational Sciences and Engineering Div.; Sulfredge, C. David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Computational Sciences and Engineering Div.

2012-12-01

This study assesses the effectiveness of block valve closure swiftness in mitigating the consequences of natural gas and hazardous liquid pipeline releases on public and environmental safety. It also evaluates the technical, operational, and economic feasibility and potential cost benefits of installing automatic shutoff valves (ASVs) and remote control valves (RCVs) in newly constructed and fully replaced transmission lines. Risk analyses of hypothetical pipeline release scenarios are used as the basis for assessing: (1) fire damage to buildings and property in Class 1, Class 2, Class 3, and Class 4 high consequence areas (HCAs) caused by natural gas pipeline releases and subsequent ignition of the released natural gas; (2) fire damage to buildings and property in HCAs designated as high population areas and other populated areas caused by hazardous liquid pipeline releases and subsequent ignition of the released propane; and (3) socioeconomic and environmental damage in HCAs caused by hazardous liquid pipeline releases of crude oil. These risk analyses use engineering principles and fire science practices to characterize thermal radiation effects on buildings and humans and to quantify the total damage cost of socioeconomic and environmental impacts. The risk analysis approach used for natural gas pipelines is consistent with risk assessment standards developed by industry and incorporated into Federal pipeline safety regulations. Feasibility evaluations for the hypothetical pipeline release scenarios considered in this study show that installation of ASVs and RCVs in newly constructed and fully replaced natural gas and hazardous liquid pipelines is technically, operationally, and economically feasible with a positive cost benefit. However, these results may not apply to all newly constructed and fully replaced pipelines because site-specific parameters that influence risk analyses and feasibility evaluations often vary significantly from one pipeline segment to

Transcatheter aortic valve-in-valve treatment of degenerative stentless supra-annular Freedom Solo valves: A single centre experience.

Science.gov (United States)

Cockburn, James; Dooley, Maureen; Parker, Jessica; Hill, Andrew; Hutchinson, Nevil; de Belder, Adam; Trivedi, Uday; Hildick-Smith, David

2017-02-15

Redo surgery for degenerative bioprosthetic aortic valves is associated with significant morbidity and mortality. Report results of valve-in-valve therapy (ViV-TAVI) in failed supra-annular stentless Freedom Solo (FS) bioprostheses, which are the highest risk for coronary occlusion. Six patients with FS valves (mean age 78.5 years, 50% males). Five had valvular restenosis (peak gradient 87.2 mm Hg, valve area 0.63 cm 2 ), one had severe regurgitation (AR). Median time to failure was 7 years. Patients were high risk (mean STS/Logistic EuroScore 10.6 15.8, respectively). FS valves ranged from 21 to 25 mm. Successful ViV-TAVI was achieved in 4/6 patients (67%). Of the unsuccessful cases, (patient 1 and 2 of series) patient 1 underwent BAV with simultaneous aortography which revealed left main stem occlusion. The procedure was stopped and the patient went forward for repeat surgery. Patient 2 underwent successful ViV-TAVI with a 26-mm CoreValve with a guide catheter in the left main, but on removal coronary obstruction occurred, necessitating valve snaring into the aorta. Among the successful cases, (patients 3, 4, 5, 6) the TAVIs used were CoreValve Evolut R 23 mm (n = 3), and Lotus 23 mm (n = 1). In the successful cases the peak gradient fell from 83.0 to 38.3 mm Hg. No patient was left with >1+ AR. One patient had a stroke on Day 2, with full neurological recovery. Two patients underwent semi-elective pacing for LBBB and PR >280 ms. ViV-TAVI in stentless Freedom Solo valves is high risk. The risk of coronary occlusion is high. The smallest possible prosthesis (1:1 sizing) should be used, and strategies to protect the coronary vessels must be considered. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

A study on the optimization of test interval for check valves of Ulchin Unit 3 using the risk-informed in-service testing approach

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.; Yang, Z. A.; Ha, J. J.

2002-01-01

We optimized the test interval for check valves of Ulchin Unit 3 using the risk-informed in-service testing (IST) approach. First, we categorized the IST check valves for Ulchin Unit 3 according to their contributions to the safety of Ulchin Unit 3. Next, we performed the risk analysis on the relaxation of test interval for check valves identified as relatively low important to the safety of Ulchin Unit 3 to identify the maximum increasable test interval of them. Finally, we estimated the number of tests of IST check valves to be performed due to the changes of test interval. These study results are as follows: The categorization of IST check valve importance; the number of the HSSCs is 24(11.48%), the ISSCs is 40 (19.14%), and the LSSCs is 462(69.38%). The maximum increasable test interval; 6 times of current test interval of ISSCs2 and 40 times of that of LSSCs. The number of tests of IST check valves to be performed during 6 refueling time can be reduced from 7692 to 1333 ( 82.7%)

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

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.

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 ...

Spatial distribution of cavitation-shock-pressure around a jet-flow gate-valve

International Nuclear Information System (INIS)

Oba, Risaburo; Takayama, Kazuyoshi; Ito, Yukio; Miyakura, Hideto; Nozaki, Satoru; Ishige, Tadashi; Sonoda, Shuji; Sakamoto, Kenji.

1987-01-01

To make clear the mechanism of cavitation erosion, the spatial distribution of cavitation shock pressures were quantitatively measured by a pressure sensitive sheet in the 1/10 scale model of a jet-flow gate-valve, for various valve-openings and cavitation numbers. The dynamic pressure response of the sheet was corrected by the shock wave generated from detonation explosives. It is made clear that the erosive shock pressures are distributed in a limited part of the whole cavitation region, and the safety region without the fatal cavitation erosion is defined. (author)

Evaluation of a coolant injection into the in-vessel with a RCS depressurization by using SCDAP/RELAP5

International Nuclear Information System (INIS)

Rae-Joon, Park; Sang-Baik, Kim; Hee-Dong, Kim

2007-01-01

As part of the evaluations of a severe accident management strategy, a coolant injection in the vessel with a reactor coolant system (RCS) depressurization has been evaluated by using the SCDAP/RELAP5 computer code. Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feed water (LOFW) accident have been analyzed in optimized power reactor OPR-1000. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 seconds with a RCS depressurization by using one condenser dump valve at 6 minutes after an entrance of the severe accident management guidance prevents a reactor vessel failure for the small break LOCA without SI. In this case, only train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent a reactor vessel failure. Only one train operation of the HPSI at 20,208 seconds with a RCS depressurization by using two safety depressurization system valves at 40 minutes after an initial opening of the safety relief valve prevents a reactor vessel failure for the total LOFW. (authors)

Cost-Effectiveness of Transcatheter Aortic Valve Replacement With a Self-Expanding Prosthesis Versus Surgical Aortic Valve Replacement.

Science.gov (United States)

Reynolds, Matthew R; Lei, Yang; Wang, Kaijun; Chinnakondepalli, Khaja; Vilain, Katherine A; Magnuson, Elizabeth A; Galper, Benjamin Z; Meduri, Christopher U; Arnold, Suzanne V; Baron, Suzanne J; Reardon, Michael J; Adams, David H; Popma, Jeffrey J; Cohen, David J

2016-01-05

Previous studies of the cost-effectiveness of transcatheter aortic valve replacement (TAVR) have been based primarily on a single balloon-expandable system. The goal of this study was to evaluate the cost-effectiveness of TAVR with a self-expanding prosthesis compared with surgical aortic valve replacement (SAVR) for patients with severe aortic stenosis and high surgical risk. We performed a formal economic analysis on the basis of individual, patient-level data from the CoreValve U.S. High Risk Pivotal Trial. Empirical data regarding survival and quality of life over 2 years, and medical resource use and hospital costs through 12 months were used to project life expectancy, quality-adjusted life expectancy, and lifetime medical costs in order to estimate the incremental cost-effectiveness of TAVR versus SAVR from a U.S. Relative to SAVR, TAVR reduced initial length of stay an average of 4.4 days, decreased the need for rehabilitation services at discharge, and resulted in superior 1-month quality of life. Index admission and projected lifetime costs were higher with TAVR than with SAVR (differences $11,260 and $17,849 per patient, respectively), whereas TAVR was projected to provide a lifetime gain of 0.32 quality-adjusted life-years ([QALY]; 0.41 LY) with 3% discounting. Lifetime incremental cost-effectiveness ratios were $55,090 per QALY gained and $43,114 per LY gained. Sensitivity analyses indicated that a reduction in the initial cost of TAVR by ∼$1,650 would lead to an incremental cost-effectiveness ratio <$50,000/QALY gained. In a high-risk clinical trial population, TAVR with a self-expanding prosthesis provided meaningful clinical benefits compared with SAVR, with incremental costs considered acceptable by current U.S. With expected modest reductions in the cost of index TAVR admissions, the value of TAVR compared with SAVR in this patient population would become high. (Safety and Efficacy Study of the Medtronic CoreValve System in the Treatment of

Consequences of a hypothetical incident for different sectors

CERN Document Server

Bertinelli, F; Garion, C; Jimenez, J M; Parma, V; Perin, A; Schmidt, R; Tavian, L; Tock, J P; van Weelderen, R

2011-01-01

During the 2009 long shutdown, the LHC machine has been partially consolidated by adding safety relief devices in order to better protect the cryostats against large helium release and consequently to mitigate the risks of collateral damages. After recalling the present relief valve implementation and other mitigations related to the collateral damages, this paper describes the damage process of a hypothetical incident, presents its consequences for the different sectors and for beam energies up to 5 TeV with emphasis on the induced downtime.

28 CFR 36.504 - Relief.

Science.gov (United States)

2010-07-01

... COMMERCIAL FACILITIES Enforcement § 36.504 Relief. (a) Authority of court. In a civil action under § 36.503, the court— (1) May grant any equitable relief that such court considers to be appropriate, including... disabilities; (2) May award other relief as the court considers to be appropriate, including monetary damages...

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 ...

Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

Energy Technology Data Exchange (ETDEWEB)

Seniuk, P.J.

1996-12-01

This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

International Nuclear Information System (INIS)

Seniuk, P.J.

1996-01-01

This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, open-quotes Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plantsclose quotes. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O ampersand M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O ampersand M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion

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...

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.)

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...

Safety improvement of Paks nuclear power plant

International Nuclear Information System (INIS)

Vamos, G.

1999-01-01

Safety upgrading completed in the early nineties at the Paks NPP include: replacement of steam generator safety valves and control valves; reliability improvement of the electrical supply system; modification of protection logic; enhancement of the fire protection; construction of full scope Training Simulator. Design safety upgrading measures achieved in recent years were concerned with: relocation of steam generator emergency feed-water supply; emergency gas removal from the primary coolant system; hydrogen management in the containment; protection against sumps; preventing of emergency core cooling system tanks from refilling. Increasing seismic resistance, containment assessment, refurbishment of reactor protection system, improving reliability of emergency electrical supply, analysis of internal hazards are now being implemented. Safety upgrading measures which are being prepared include: bleed and feed procedures; reactor over-pressurisation protection in cold state; treatment of steam generator primary to secondary leak accidents. Operational safety improvements are dealing with safety culture, training measures and facilities; symptom based emergency operating procedures; in-service inspection; fire protection. The significance of international cooperation is emphasised in view of achieving nuclear safety standards recognised in EU

Gasoline New Timing and Flux Adjustable Rotary Valve Design (Hereinafter: Rotary Valve

Directory of Open Access Journals (Sweden)

Du huiqi

2016-01-01

Full Text Available Conventional gasoline engine with an umbrella valve control cylinder intake and exhaust, in order to achieve sealing effect, the valve is driven by the spring force; at the same time, when the cam opens the valve to overcome the spring force acting. Sealing the better, the more power consumed in the engine mechanical losses, the valve mechanism consumes about 30%, which is not a small loss! This article describes a new type of rotary valve is to significantly reduce mechanical losses, so as to achieve energy saving purposes.

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)

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 ...

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 ...

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.

Motor-operated Valve Program at NPP Krsko (NEK) - Status and Overview

International Nuclear Information System (INIS)

Nikolic, M; Jagodar, N.; Cerjak, J.; Butkovic, V.

2002-01-01

On the basics of US NRC Generic Letter 89-10 Safety-related Motor-operated Valve Testing and Surveillance and subsequent generic letters, Motor-operated Valve (MOV) Program at NEK has been developing. Namely, the holders of nuclear power plant operating licenses has to verify the design basis capability of safety-related e.g. important-to-safety MOVs, as well as to ensure the same for the life of the plant. In light of that, each plant should establish a program to address stressed issues for each program MOV (124 at NEK). Such comprehensive task requires significant effort in many aspects, and basically multidisciplinary skills. NEK MOV Program represents a blend of engineering and in-plant testing, comprised of three phases: Phase I Engineering, Phase II Field Implementation and Phase III Trending. Currently, the program is about the end of Phase I and II, as well as in development of engineering basis for launching Phase III. Overview of the major programmatic issues will be given in this paper along with ongoing activities: testing process, gear-ratio modification, pressure locking/thermal binding susceptibility screening and preventive maintenance. (author)

Mid-term study of transcatheter aortic valve implantation in an Asian population with severe aortic stenosis: two-year Valve Academic Research Consortium-2 outcomes.

Science.gov (United States)

Chew, Nicholas; Hon, Jimmy Kim Fatt; Yip, Wei Luen James; Chan, Siew Pang; Poh, Kian-Keong; Kong, William Kok-Fai; Teoh, Kristine Leok Kheng; Yeo, Tiong Cheng; Tan, Huay Cheem; Tay, Edgar Lik Wui

2017-09-01

Transcatheter aortic valve implantation (TAVI) is an effective treatment for high-risk or inoperative patients with severe aortic stenosis. Given the unique characteristics of Asian populations, questions regarding mid-term outcomes in Asians undergoing TAVI have yet to be addressed. We evaluated the two-year clinical outcomes of TAVI in an Asian population using Valve Academic Research Consortium-2 definitions. This prospective study recruited 59 patients from a major academic medical centre in Singapore. The main outcomes were two-year survival rates, peri-procedural complications, symptom improvement, valvular function and assessment of learning curve. Mean age was 76.8 years (61.0% male), mean body surface area 1.6 m 2 and mean logistic EuroSCORE 18.7%. Survival was 93.2%, 86.0% and 79.1% at 30 days, one year and two years, respectively. At 30 days post TAVI, the rate of stroke was 1.7%, life-threatening bleeding 5.1%, acute kidney injury 25.0%, major vascular complication 5.1%, and new permanent pacemaker implantation 6.8%. 29.3% of TAVI patients were rehospitalised (47.1% cardiovascular-related) within one year. These composite outcomes were measured: device success (93.2%); early safety (79.7%); clinical efficacy (66.1%); and time-related valve safety (84.7%). Univariate analysis found these predictors of two-year all-cause mortality: logistic EuroSCORE (hazard ratio [HR] 1.07; p < 0.001); baseline estimated glomerular filtration rate (HR 0.97; p = 0.048); and acute kidney injury (HR 5.33; p = 0.022). Multivariate analysis identified non-transfemoral TAVI as a predictor of cardiovascular-related two-year mortality (HR 14.64; p = 0.008). Despite the unique clinical differences in Asian populations, this registry demonstrated favourable mid-term clinical and safety outcomes in Asians undergoing TAVI. Copyright: © Singapore Medical Association

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.

Transapical JenaValve in a patient with mechanical mitral valve prosthesis.

LENUS (Irish Health Repository)

O' Sullivan, Katie E

2014-01-29

We report the first case of transcatheter aortic valve replacement implantation using JenaValve™ in a patient with mechanical mitral valve prosthesis. We believe that the design features of this valve may be particularly suited for use in this setting. © 2014 Wiley Periodicals, Inc.

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.)

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.

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.

Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

Presentation, management, and outcome of posterior urethral valves in a Nigerian tertiary hospital

Directory of Open Access Journals (Sweden)

Agbugui Jude Orumuah

2015-01-01

Full Text Available Background: Posterior urethral valves (PUV remain the most common cause of bladder outlet obstruction and renal insufficiency in male children. The aim of this study was to evaluate the presentation, management, challenges, and outcome of the disease in a Nigerian tertiary health institution. Patients and Methods: Retrospectively, medical records of male children with a diagnosis of PUVs over a 10 year period (2003-2012 were retrieved. All data in relation to the study objectives were recorded and analyzed. Results: A total of 44 patients was managed for PUV within the period. The mean age of presentation was 3.95 years with 56.8% of the patients presenting after the age of 1 year. Voiding dysfunction noted in 40 (91.0% patients was the most common mode of presentation. The most common finding on physical examination was a palpable bladder while urinary tract infection noted in 23 (52.3% patients was the most common complication noted. Abdominal ultrasonography revealed dilated posterior urethra in 16 (36.4% cases, while micturating cystourethrogram revealed a dilated proximal urethra in all 35 cases in which it was done, diverticulum in 6 and vesicoureteric reflux in 9. The creatinine value at presentation ranged between 0.4 mg/dl and 4.0 mg/dl with a mean of 1.02 ± 0.93 mg/dl. Urethroscopy in 37 patients confirmed type I and type III PUV in 35 and 2 patients, respectively. Valve ablation with a diathermy bugbee electrode provided relief of obstructions in the 37 patients who underwent the procedure without any significant immediate complication. The period of follow-up ranged between 2 weeks and 3 years with a mean of 10.2 months. There was sustained improvement in urine stream, reduction in the mean creatinine concentration and incidence of UTI during follow-up. However, patients with significantly impaired renal function had a poorer outcome. Conclusion: Many patients with PUV presented late within the reviewed period. Valve ablation

Transaortic edge-to-edge mitral valve repair for moderate secondary/functional mitral regurgitation in patients undergoing aortic root/valve intervention.

Science.gov (United States)

Choudhary, Shiv Kumar; Abraham, Atul; Bhoje, Amol; Gharde, Parag; Sahu, Manoj; Talwar, Sachin; Airan, Balram

2017-11-01

The present study evaluates the feasibility, safety, and efficacy of edge-to-edge repair for moderate secondary/functional mitral regurgitation in patients undergoing aortic valve/root interventions. Sixteen patients underwent transaortic edge-to-edge mitral valve repair. Mitral regurgitation was 2+ in 8 patients and 3+ in 6 patients. Two patients in whom cardiac arrest developed preoperatively had severe (4+) mitral regurgitation. Patients underwent operation for severe aortic regurgitation ± aortic root lesions. The mean left ventricular systolic and diastolic diameters were 51.5 ± 12.8 mm and 70.7 ± 10.7 mm, respectively. Left ventricular ejection fraction ranged from 20% to 60%. Primary surgical procedure included Bentall's ± hemiarch replacement in 10 patients, aortic valve replacement in 5 patients, and noncoronary sinus replacement with aortic valve repair in 1 patient. Severity of mitral regurgitation decreased to trivial or zero in 13 patients, 1+ in 2 patients, and 2+ in 1 patient. There were no gradients across the mitral valve in 9 patients, less than 5 mm Hg in 6 patients, and 9 mm Hg in 1 patient. There was no operative mortality. Follow-up ranged from 2 weeks to 54 months. Echocardiography showed trivial or no mitral regurgitation in 12 patients, 1+ in 2 patients, and 2+ in 2 patients. None of the patients had significant mitral stenosis. The mean left ventricular systolic and diastolic diameters decreased to 40.5 ± 10.3 mm and 58.7 ± 11.6 mm, respectively. Ejection fraction also improved slightly (22%-65%). Transaortic edge-to-edge mitral valve repair is a safe and effective technique to abolish secondary/functional mitral regurgitation. However, its impact on overall survival needs to be studied. Copyright © 2017 The American Association for Thoracic Surgery. Published by Elsevier Inc. All rights reserved.

Use of a valve operation test and evaluation system to enhance valve reliability

International Nuclear Information System (INIS)

Lowry, D.A.

1990-01-01

Power plant owners have emphasized the need for assuring safe, reliable operation of valves. While most valves must simply open or close, the mechanisms involved can be quite complex. Motor operated valves (MOVs) must be properly adjusted to assure operability. Individual operator components determine the performance of the entire MOV. Failure in MOVs could cripple or shut down a unit. Thus, a complete valve program consisting of design reviews, operational testing, and preventive and predictive maintenance activities will enhance an owner's confidence level that his valves win operate as expected. Liberty's Valve Operation Test and Evaluation System (VOTES) accurately measures stein thrust without intruding on valve operation. Since mounting a strain gage to a valve stem is a desirable but impractical way of obtaining precise stem thrust, Liberty developed a method to obtain identical data by placing a strain gage sensor on the valve yoke. VOTES provides information which effectively eliminates costly, unscheduled downtime. This paper presents the results of infield VOTES testing. The system's proven ability to identify and characterize actuator and valve performance is demonstrated. Specific topics of discussion include the ability of VOTES to ease a utility's IE Bulletin 8543 concerns and conclusively diagnose MOV components. Data from static and differential pressure testing are presented. Technical, operational, and financial advantages resulting from VOTES technology are explored in detail

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

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

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.

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

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…