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

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...

Refuge alternatives relief valve testing and design with updated test stand.

Science.gov (United States)

Lutz, T J; Bissert, P T; Homce, G T; Yonkey, J A

2018-03-01

Underground refuge alternatives require an air source to supply breathable air to the occupants. This requires pressure relief valves to prevent unsafe pressures from building up within the refuge alternative. The U.S. Mine Safety and Health Administration (MSHA) mandates that pressure relief valves prevent pressure from exceeding 1.25 kPa (0.18 psi), or as specified by the manufacturer, above mine atmospheric pressure when a fan or compressor is used for the air supply. The U.S. National Institute for Occupational Safety and Health (NIOSH) tested a variety of pressure relief valves using an instrumented test fixture consisting of data acquisition equipment, a centrifugal blower, ductwork and various sensors to determine if the subject pressure relief valves meet the MSHA requirement. Relief pressures and flow characteristics, including opening pressure and flow rate, were measured for five different pressure relief valves under a variety of conditions. The subject pressure relief valves included two off-the-shelf modified check valves, two check valves used in MSHA-approved built-in-place refuge alternatives, and a commercially available valve that was designed for a steel refuge alternative and is currently being used in some built-in-place refuge alternatives. The test results showed relief pressures ranging from 0.20 to 1.53 kPa (0.03 to 0.22 psi) and flow rates up to 19.3 m 3 /min (683 scfm). As tested, some of the pressure relief valves did not meet the 1.25 kPa (0.18 psi) relief specification.

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.

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.

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.)

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...

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

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)

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

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

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

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.)

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

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

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

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

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

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)

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

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 relief valve alternate analysis method

International Nuclear Information System (INIS)

Adams, R.H.; Javid, A.; Khatua, T.P.

1981-01-01

An experimental test program was started in the United States in 1976 to define and quantify Safety Relief Valve (SRV) phenomena in General Electric Mark I Suppression Chambers. The testing considered several discharged devices and was used to correlate SRV load prediction models. The program was funded by utilities with Mark I containments and has resulted in a detailed SRV load definition as a portion of the Mark I containment program Load Definition Report (LDR). The (USNRC) has reviewed and approved the LDR SRV load definition. In addition, the USNRC has permitted calibration of structural models used for predicting torus response to SRV loads. Model calibration is subject to confirmatory in-plant testing. The SRV methodology given in the LDR requires that transient dynamic pressures be applied to a torus structural model that includes a fluid added mass matrix. Preliminary evaluations of torus response have indicated order of magnitude conservatisms, with respect to test results, which could result in unrealistic containment modifications. In addition, structural response trends observed in full-scale tests between cold pipe, first valve actuation and hot pipe, subsequent valve actuation conditions have not been duplicated using current analysis methods. It was suggested by others that an energy approach using current fluid models be utilized to define loads. An alternate SRV analysis method is defined to correct suppression chamber structural response to a level that permits economical but conservative design. Simple analogs are developed for the purpose of correcting the analytical response obtained from LDR analysis methods. Analogs evaluated considered forced vibration and free vibration structural response. The corrected response correlated well with in-plant test response. The correlation of the analytical model at test conditions permits application of the alternate analysis method at design conditions. (orig./HP)

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%)

Recent performance experience with US light water reactor self-actuating safety and relief valves

Energy Technology Data Exchange (ETDEWEB)

Hammer, C.G.

1996-12-01

Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

Fairchild Stratos Division's Type II prototype lockhopper valve: METC Prototype Test Valve No. F-1 prototype lockhopper valve-testing and development project. Static test report

Energy Technology Data Exchange (ETDEWEB)

Goff, D. R.; Cutright, R. L.; Griffith, R. A.; Loomis, R. B.; Maxfield, D. A.; Moritz, R. S.

1981-10-01

METC Prototype Test Valve No. F-1 is a hybrid design, based on a segmented ball termed a visor valve, developed and manufactured by Fairchild Stratos Division under contract to the Department of Energy. The valve uses a visor arm that rotates into position and then translates to seal. This valve conditionally completed static testing at METC with clean gas to pressures of 1600 psig and internal valve temperatures to 600/sup 0/F. External leakage was excessive due to leakage through the stuffing box, purge fittings, external bolts, and other assemblies. The stuffing box was repacked several times and redesigned midway through the testing, but external leakage was still excessive. Internal leakage through the seats, except for a few anomalies, was very low throughout the 2409 cycles of testing. As shown by the low internal leakage, the visor valve concept appears to have potential for lock-hopper valve applications. The problems that are present with METC Prototype Test Valve No. F-1 are in the seals, which are equivalent to the shaft and bonnet seals in standard valve designs. The operating conditions at these seals are well within the capabilities of available seal designs and materials. Further engineering and minor modifications should be able to resolve the problems identified during static testing.

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

Trend analysis of incidents involving setpoint drift in safety or safety/relief valves at U.S. LWRs

International Nuclear Information System (INIS)

Watanabe, Norio

2008-01-01

Since the beginning of the 1980's, in the United States, there have been many licensee event reports (LERs) involving setpoint drift in safety or safety/relief valves. The United States Nuclear Regulatory Commission (NRC) has issued a lot of generic communications on this issue and the industry has made efforts to resolve the issue. However, the NRC staff recently highlighted that over 70 LERs involved instances where safety or safety/relief valves failed to meet the allowed setpoint tolerance from 2001 through August 2006. In the present study, we analyzed the U.S. experience with setpoint drift in safety/relief valves (SRVs) at BWRs, pressurizer safety valves (PSVs), and main steam safety valves (MSSVs) at PWRs by reviewing approximately 90 LERs from 2000 to 2006 and examined the trend focusing on causes and setpoint deviation ranges. This study indicates that for SRVs and MSSVs, disc-seat bonding is a dominant cause of the setpoint drifting high and has a tendency to result in a relatively large deviation of the setpoint. This means that disc-seat bonding might be a safety concern from the view point of overpressure protection. For PSVs, the deviation of setpoints is generally small, although its causes are not specified in many instances. (author)

Structural analysis strategies of the pressurized relief and safety valves discharge piping of NPP Angra 1

International Nuclear Information System (INIS)

Lima, Maria Ines Prates de; Kuramoto, Edson; Suanno, Rodolfo

2002-01-01

The pressurizer relief and safety valve system provides the reactor coolant system overpressure protection and, therefore, it is fundamental for the security of a nuclear plant. This paper discusses the safety valve loop seal strategies adopted by others nuclear power plants over the world in order to attend the recommendations of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations). The technical option adopted for Angra 1 consists in making specific modifications on the original piping and support configuration of the pressurizer relief and safety valve system. These modifications were proposed in order to reduce the high stress levels induced by the thermal-hydrodynamic loads caused by the discharge of the sub-cooled water during the opening of the relief or the safety valves. Several thermal-hydraulic models were tested to assess the influence of the seal water heating and the simultaneous opening of the valves in order to minimize the thermal hydrodynamic loads effects. The piping structural analysis was performed, using the computer program system KWUROHR, to satisfy the requirements of the appropriate equations of the code ASME Section III, Subsections NB3650 and NC3650. (author)

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

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 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

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.

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.

Structural concept of angle type of hot isolation valve and its test program at an out-of-pile test facility

Energy Technology Data Exchange (ETDEWEB)

Hada, Kazuhiko; Fujisaki, Katsuo; Shibata, Taijyu; Inagaki, Yoshiyuki; Hino, Ryutaro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Koiso, Hiroshi

1997-02-01

The Japanese safety regulation generally requires to set an isolation valve at the penetration of the reactor containment vessel on the secondary helium piping system which connects a steam reforming hydrogen production system, located outside the reactor building, to an intermediate heat exchanger (IHX) in the HTTR reactor system. The hot secondary helium which is heated up to the high temperature of 905degC and at the high pressure of 4.1MPa is passing through the isolation valve. So far, such a hot isolation valve has not been industrialized. The present report presents a proposal of a structural design concept of an angle valve as a promising candidate of the hot isolation valve, and a proposal on a test program for demonstrating the technological feasibility of the concept at an out-of-pile test facility before installing at the HTTR. A closing time and a leak rate at a valve seat are the key design parameters for developing the design concept. To set a reasonable value to each parameter, safety requirements on the isolation valve were discussed at first. The target closing time and the acceptable design limit of leak rate at the valve seat for meeting the requirements were specified 30 seconds and 10 STP cm{sup 3}/s, respectively. A nickel-base superalloy Hastelloy XR is feasible as such a valve seat material as to withstand the internal/external pressure of 4.1MPa at the high temperature of 905degC, the severest loading conditions of the valve seat at the accident of secondary helium pipe rupture. Correlation of leak rate at the ambient temperature to that at an operating temperature (900degC) is one of key test subjects of test program at an out-of-pile test facility. Leak rate at the operating temperature is the real parameter to be checked but only the leak rate at the ambient temperature is measured at regulatory examination in service. A test method to develop such correlation was proposed. (author)

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.

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.)

Cyclonic valve test: preliminary results

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

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

Main steam system piping response under safety/relief valve opening events

International Nuclear Information System (INIS)

Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

1980-01-01

The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

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

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.

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

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

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.

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.

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.

Motor operated valves problems tests and simulations

Energy Technology Data Exchange (ETDEWEB)

Pinier, D.; Haas, J.L.

1996-12-01

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

Motor operated valves problems tests and simulations

International Nuclear Information System (INIS)

Pinier, D.; Haas, J.L.

1996-01-01

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

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

Acoustic emission testing of piston check valves

International Nuclear Information System (INIS)

Stewart, D.L.

1994-01-01

Based on test experience at Comanche Peak Unit 1, an acoustic emission data evaluation matrix for piston check valves has been developed. The degradations represented in this matrix were selected based on Edwards piston check valve failure data reported in the Nuclear Plant Reliability Data System. Evidence to support this matrix was collected from site test data on a variety of valve types. Although still under refinement, the matrix provides three major attributes for closure verification, which have proven useful in developing test procedures for inservice testing and preventing unnecessary disassembly

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)

Light water reactor pressure isolation valve performance testing

International Nuclear Information System (INIS)

Neely, H.H.; Jeanmougin, N.M.; Corugedo, J.J.

1990-07-01

The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (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...

Routine testing on protective and safety systems and components

International Nuclear Information System (INIS)

Rysy, W.

1977-01-01

1) In-process inspection, tests during commissioning. 2) Tests during reactor operation. 2.1) Reactor protection system, for example: continuous auto-testing by a dynamic system, check of the output signals; 2.2) safety features: selected examples: functional tests on the ECCS, trial operation of the emergency diesels. 3) Tests during refuelling phase. 3.1) Containment: Leakage rate tests, leak testing; 3.2) coolant system: selected examples: inservice inspections of the pressure vessel, eddy current testing of the steam generator, functional tests of safety valves. (orig./HP) [de

Technical evaluation report, TMI action NUREG-0737 (II.D.1), relief and safety valve testing, Comanche Peak, Unit 2, Docket No. 50-446

International Nuclear Information System (INIS)

Fineman, C.P.

1993-01-01

In the past, safety and relief valves installed in the primary coolant system of light water reactors have performed improperly. As a result, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) and, subsequently, NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended development and completion of programs to do two things. First, the programs should reevaluate the functional performance capabilities of pressurized water reactor safety, relief, and block valves. Second, they should verify the integrity of the pressurizer safety and relief valve piping systems for normal, transient, and accident conditions. This report documents the review of those programs by EG ampersand G Idaho, Inc. Specifically, this report documents the review of the Comanche Peak, Unit 2, Applicant response to the requirement of NUREG-0578 and NUREG-0737. This review found the Applicant provided an acceptable response reconfirming they met General Design Criteria 14, 15 and 30 of Appendix A to 10 CFR 50 for the subject equipment

Fast Flux Test Facility primary sodium valves

International Nuclear Information System (INIS)

Rabe, G.B.; Ezra, B.C.

1977-01-01

The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

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...

Prototype bellows sealed nuclear valve development -reliability through testing

International Nuclear Information System (INIS)

Dixon, D.F.; Abbas, M.

1978-01-01

To assist in appraising bellows sealed valve performance, 10 tests were done on a ''1 in.'' prototype bellows sealed valve design. The tests simulated primary heat transport (PHT) system conditions for a 600 MWe CANDU-PHW. The design approach was to have all valve components outlast the bellows in endurance tests; this was achieved. The valve design meets the Atomic Energy of Canada Limited specification. For comparison, bellows fatigue failure data were fitted to both log-normal and Weibull distributions. A numerical example shows how to select valve stroke amplitude on the basis of valve flow requirement and the minimum acceptable fatigue life. (author)

The need to optimize inservice testing and inspection to enhance safety

International Nuclear Information System (INIS)

Perry, J.A.

1996-01-01

Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view

The need to optimize inservice testing and inspection to enhance safety

Energy Technology Data Exchange (ETDEWEB)

Perry, J.A.

1996-12-01

Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view.

Durability Tests of Ball Valve Prototype with Flowmeter Operation

Science.gov (United States)

Rogula, J.; Romanik, G.

2018-02-01

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

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

Environmental qualification testing of TFE valve components

International Nuclear Information System (INIS)

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

1997-01-01

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

Early results of gate valve flow interruption blowdown tests

International Nuclear Information System (INIS)

DeWall, K.G.

1988-01-01

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

NRC Information No. 90-18: Potential problems with Crosby safety relief valves used on diesel generator air start receiver tanks

International Nuclear Information System (INIS)

Rossi, C.E.

1992-01-01

On March 31, 1989, Cooper Industries was made aware of circumstances at Perry Unit 1 that led to the Division I EDG being declared inoperable. A Crosby safety relief valve on one of the two EDG starting air receiving tanks was inadvertently hit during maintenance activities. The force of the impact caused the valve to open and blow down both air receiving tanks. The safety relief valve did not reseat until approximately 30 psig below the EDG automatic start lockout signal. On January 12, 1990, Cooper Industries learned that a similar event had occurred at Comanche Peak. On January 17, 1990, Cooper Industries submitted a 10 CFR Part 21 report on the affected safety relief valves (Crosby style JMBU and JRU safety relief valves). Although Crosby-style JMBU and JRU safety relief valves were designed to meet the requirements of Section VIII of the ASME Boiler and Pressure Vessel Code, they were not seismically qualified. In addition, the blowdown characteristics of the valves were not consistent with the functional requirements of the system in which they were installed. Cooper Industries has recommended replacing these valves with seismically qualified valves that have the proper blowdown reseat characteristics

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)

Reliability Evaluation of Concentric Butterfly Valve Using Statistical Hypothesis Test

Energy Technology Data Exchange (ETDEWEB)

Chang, Mu Seong; Choi, Jong Sik; Choi, Byung Oh; Kim, Do Sik [Korea Institute of Machinery and Materials, Daejeon (Korea, Republic of)

2015-12-15

A butterfly valve is a type of flow-control device typically used to regulate a fluid flow. This paper presents an estimation of the shape parameter of the Weibull distribution, characteristic life, and B10 life for a concentric butterfly valve based on a statistical analysis of the reliability test data taken before and after the valve improvement. The difference in the shape and scale parameters between the existing and improved valves is reviewed using a statistical hypothesis test. The test results indicate that the shape parameter of the improved valve is similar to that of the existing valve, and that the scale parameter of the improved valve is found to have increased. These analysis results are particularly useful for a reliability qualification test and the determination of the service life cycles.

Reliability Evaluation of Concentric Butterfly Valve Using Statistical Hypothesis Test

International Nuclear Information System (INIS)

Chang, Mu Seong; Choi, Jong Sik; Choi, Byung Oh; Kim, Do Sik

2015-01-01

A butterfly valve is a type of flow-control device typically used to regulate a fluid flow. This paper presents an estimation of the shape parameter of the Weibull distribution, characteristic life, and B10 life for a concentric butterfly valve based on a statistical analysis of the reliability test data taken before and after the valve improvement. The difference in the shape and scale parameters between the existing and improved valves is reviewed using a statistical hypothesis test. The test results indicate that the shape parameter of the improved valve is similar to that of the existing valve, and that the scale parameter of the improved valve is found to have increased. These analysis results are particularly useful for a reliability qualification test and the determination of the service life cycles

Studies and testing in water and steam of valves and fittings, and nuclear components. The result of 25 years of testing using a comprehensive range of test facilities under service conditions

International Nuclear Information System (INIS)

Berail, J.F.; Bruneau, S.; Crouzet, D.; Haas, J.L.; Zbinden, M.

1998-05-01

Electricite de France operates 58 PWR nuclear power stations, for which the behaviour of valves and fittings is of major importance for safety, for the availability of the plants, and for maintenance costs. Since the early 70's, EDF has developed a comprehensive range of facilities to test valves and fittings in PWR service and accident conditions. It has carried out studies, tests, development work, experimental and numerical research in collaboration with external organisations and manufacturers, to improve the technologies of these equipment as well as maintenance tools and methods. In the present paper, the authors quantify the importance of valves and fittings studies for EDF, which has led to the drawing up of a catalogue of approved equipment. They describe the principle test facilities, and the structure of the EDF 'valves and fittings tests results' data base. They show the importance of twenty-five years of testing experience for both the evolution of equipment and for the increase in French nuclear plants availability. (author)

Testing of cobalt-free alloys for valve applications using a special test loop

International Nuclear Information System (INIS)

Benhamou, C.

1992-01-01

Considering that use of cobalt alloys should be avoided as far as possible in PWR components, a programme aimed at establishing the performance of cobalt-free alloys has been performed for valve applications, where cobalt alloys are mainly used. Referring to past work, two types of cobalt-free alloys were selected: Ni-Cr-B-Si and Ni-Cr-Fe alloys. Cobalt-free valves' behaviour has been evaluated comparatively with cobalt valves by implementation of a programme in a special PWR test loop. At the issue of the loop test programme, which included endurance, thermal shock and erosion tests, cobalt-free alloys candidate to replace cobalt alloys are proposed in relation with valve type (globe valve and swing check valve). The following was established: (i) Colmonoy 4-26 (Ni-Cr-B-Si alloy) and Cenium Z20 (Ni-Cr-Fe alloy) deposited by plasma arc process were found suitable for use in 3inch swing check valves; (ii) for integral parts acting as guide rings, Nitronic 60 and Cesium Z20/698 were tested successfully; (iii) for small-bore components such as 2inch globe valves, no solution can yet be proposed; introduction of cobalt-free alloys is dependent on the development of automatic advanced arc surfacing techniques applied to small-bore components

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

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

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.

Experimental apparatus to test air trap valves

Science.gov (United States)

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

2010-08-01

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

Experimental apparatus to test air trap valves

Energy Technology Data Exchange (ETDEWEB)

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

2010-08-15

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

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

Preliminary observations of gate valve flow interruption tests, Phase 2

International Nuclear Information System (INIS)

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

1990-01-01

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

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.

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

A Methodology for Evaluation of Inservice Test Intervals for Pumps and Motor Operated Valves

International Nuclear Information System (INIS)

McElhaney, K.L.

1999-01-01

The nuclear industry has begun efforts to reevaluate inservice tests (ISTs) for key components such as pumps and valves. At issue are two important questions--What kinds of tests provide the most meaningful information about component health, and what periodic test intervals are appropriate? In the past, requirements for component testing were prescribed by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The tests and test intervals specified in the Code were generic in nature and test intervals were relatively short. Operating experience has shown, however, that performance and safety improvements and cost savings could be realized by tailoring IST programs to similar components with comparable safety importance and service conditions. In many cases, test intervals may be lengthened, resulting in cost savings for utilities and their customers

Test-circuits for HVDC thyristor valves

NARCIS (Netherlands)

Thiele, G.; Alarovandi, G.; Bonfanti, I.; Cepek, M.; Dumrese, H.G.; Damstra, G.C.

1997-01-01

In conformity with the usual classification of the specified tests into two major categories, dielectric tests and operational tests, the report is in two parts : Part 1 - Dielectric tests which deals principally with test circuits for verifying the high voltage characteristics of the valve, and

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)

Relief valve testing study

International Nuclear Information System (INIS)

BROMM, R.D.

2001-01-01

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

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)

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).

Results of the motor-operated valve engineering and testing program

International Nuclear Information System (INIS)

Black, B.R.

1994-01-01

The Texas Utilities Electric Company (TU Electric) motor-operated valve (MOV) program for implementing the recommendations of Generic Letter 89-10 has typically included the following: refurbishing each actuator, verifying each actuator's as-built configuration, testing each actuator's motor on a dynamometer, testing each actuator's torque spring pack (which is used to control the torque developed), testing each fully refurbished and reassembled actuator on a torque test stand, and testing as many MOVs as practicable both without fluid flow through the valve and with the maximum test conditions reasonably achievable (static and differential pressures (DP) conditions, respectively). Test data are acquired at 1,000 samples per second for stem thrust, stem torque, stem position, actuator compensator spring pack deflection, actuator torque spring pack deflection, motor current, motor voltage, motor three-phase power, valve upstream pressure, and valve downstream pressure, wherever practicable

Steam turbine power plant having improved testing method and system for turbine inlet valves associated with downstream inlet valves preferably having feedforward position managed control

International Nuclear Information System (INIS)

Lardi, F.; Ronnen, U.G.

1981-01-01

A throttle valve test system for a large steam turbine functions in a turbine control system to provide throttle and governor valve test operations. The control system operates with a valve management capability to provide for pre-test governor valve mode transfer when desired, and it automatically generates feedforward valve position demand signals during and after valve tests to satisfy test and load control requirements and to provide smooth transition from valve test status to normal single or sequential governor valve operation. A digital computer is included in the control system to provide control and test functions in the generation of the valve position demand signals

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.

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Staunton, R.H.; Cox, D.F. [Oak Ridge National Lab., TN (United States)

1995-03-01

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail.

Aging and service wear of spring-loaded pressure relief valves used in safety-related systems at nuclear power plants

International Nuclear Information System (INIS)

Staunton, R.H.; Cox, D.F.

1995-03-01

Spring-loaded pressure relief valves (PRVS) are used in some safety-related applications at nuclear power plants. In general, they are used in systems where, during accidents, pressures may rise to levels where pressure safety relief is required for protection of personnel, system piping, and components. This report documents a study of PRV aging and considers the severity and causes of service wear and how it is discovered and corrected in various systems, valve sizes, etc. Provided in this report are results of the examination of the recorded failures and identification of trends and relationships/correlations in the failures when all failure-related parameters are considered. Components that comprise a typical PRV, how those components fail, when they fail, and the current testing frequencies and methods are also presented in detail

Analytical modeling of bwr safety relief valve blowdown phenomenon

International Nuclear Information System (INIS)

Hwang, J.G.; Singh, A.

1984-01-01

An analytical, qualitative understanding of the pool pressures measured during safety relief valve discharge in boiling water reactors equipped with X-quenchers has been developed and compared to experimental data. A pressure trace typically consists of a brief 25-35 Hz. oscillation followed by longer 5-15 Hz. oscillation. In order to explain the pressure response, a discharge line vent clearing model has been coupled with a Rayleigh bubble dynamic model. The local conditions inside the safety relief valve discharge lines and inside of the X-quencher were simulated successfully with RELAP5. The simulation allows one to associate the peak pressure inside the quencher arm with the onset of air discharge into the suppression pool. Using the pressure and thermodynamic quality at quencher exit of RELAP5 calculation as input, a Rayleigh model of pool bubble dynamics has successfully explained both the higher and lower frequency pressure oscillations. The higher frequency oscillations are characteristic of an air bubble emanating from a single row of quencher holes. The lower frequency pressure oscillations are characteristic of a larger air bubble containing all the air expelled from one side of an X-quencher arm

Tests for the dynamic behavior of insulation valve screws

International Nuclear Information System (INIS)

Tulke, K.D.; Stoppler, W.; Stern, G.

1994-01-01

Thermal tensile tests were performed at a temperature of 270 C, with two new original insulation valve conical screws M30-Tx92,5 mm (material: 21 CrMo V 5 7)and two prestrained ones during the event on 27.12.92. In order to assure the results obtained with regard to the dynamic load on the insulation valve during ''quick opening'', in addition tensile impact tests were performed at 270 C with six original insulation valve conical screws. Impact velocity reached 13,5 m/s at four screws and 6 m/s at two screws. Test conditions regarding collision damping and mass distribution were adapted, by means of parameter studies, to the situation of the insulation valve. During thermal tensile tests, strength and deformation values, such as stress at flow start, tensile strength, fracture prolongation and strain, necking at fracture as well as energy absorption up to maximum force and up to rupture, were determined. During tensile impact tests, deformation values, such as elongation, strain and necking, and energy absorption by the screw, were determined. (orig.) [de

Are anticoagulant independent mechanical valves within reach-fast prototype fabrication and in vitro testing of innovative bi-leaflet valve models.

Science.gov (United States)

Scotten, Lawrence N; Siegel, Rolland

2015-08-01

Exploration for causes of prosthetic valve thrombogenicity has frequently focused on forward or post-closure flow detail. In prior laboratory studies, we uncovered high amplitude flow velocities of short duration close to valve closure implying potential for substantial shear stress with subsequent initiation of blood coagulation pathways. This may be relevant to widely accepted clinical disparity between mechanical and tissue valves vis-à-vis thrombogenicity. With a series of prototype bi-leaflet mechanical valves, we attempt reduction of closure related velocities with the objective of identifying a prototype valve with thrombogenic potential similar to our tissue valve control. This iterative design approach may find application in preclinical assessment of valves for anticoagulation independence. Tested valves included: prototype mechanical bi-leaflet BVs (n=56), controls (n=2) and patented early prototype mechanicals (n=2) from other investigators. Pulsatile and quasi-steady flow systems were used for testing. Projected dynamic valve area (PDVA) was measured using previously described novel technology. Flow velocity over the open and closing periods was determined by volumetric flow rate/PDVA. For the closed valve interval, use was made of data obtained from quasi-steady back pressure/flow tests. Performance was ranked by a proposed thrombogenicity potential index (TPI) relative to tissue and mechanical control valves. Optimization of the prototype valve designs lead to a 3-D printed model (BV3D). For the mitral/aortic site, BV3D has lower TPI (1.10/1.47) relative to the control mechanical valve (3.44/3.93) and similar to the control tissue valve (ideal TPI ≤1.0). Using unique technology, rapid prototyping and thrombogenicity ranking, optimization of experimental valves for reduced thrombogenic potential was expedited and simplified. Innovative mechanical valve configurations were identified that merit consideration for further development which may bring

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

Test Rig for Valves of Digital Displacement Machines

DEFF Research Database (Denmark)

Nørgård, Christian; Christensen, Jeppe Haals; Bech, Michael Møller

2017-01-01

A test rig for the valves of digital displacement machines has been developed at Aalborg University. It is composed of a commercial radial piston machine, which has been modified to facilitate Digital Displacement operation for a single piston. Prototype valves have been optimized, designed and m...

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

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)

Simulation of Valve Operation for Flow Interrupt Test in Nuclear Power Plant

International Nuclear Information System (INIS)

Kim, Jae Hyung; Shin, Dae Yong; Shin, Dong Woo; Kim, Charn Jung; Lee, Jung Hee

2012-01-01

The valve used in nuclear power plant must be qualified for the function according to the KEPIC MF. The test valve must be selected by shape and size, which is given by KEPIC MF. In the functional test, the mathematical model for the valve operation is needed. The mathematical model must be verified by the test, whose method and procedure is defined in KEPIC MF. The lack of analytical technique has lead to the poor mathematical model, with which the functional test for the big valve is impossible with analytical method. Especially, the tank and rupture disk in the flow test is not considered and the result of the analysis is so different to the real one. In these days, the 3D model for the flow interrupt test makes more accurate analysis. And no facility about functional test reduces the research will for the nuclear power plant valve. For this problem, the test facility for the functional test of the valve and pump in nuclear power plant has been made until 2012. With the test facility, the research project related the valve were initiated in KIMM( Korea Institute of Machinery and Materials). And the joint project to SNU(Seoul National University) has been going on the numerical analysis for the valve in nuclear power plant. Using the commercial software and user subroutine, UDF, the co-simulation with multi-body dynamic and fluid flow analysis and the addition of tank and rupture disk to the user subroutine make possible to simulate the flow interrupt test numerically. This is not simple and regular analysis, which was introduced in user subroutine. In order to simulate the real situation, the engineering work, related mathematical model, and the programming in the user subroutine are needed. This study is on the making the mathematical model for the functional test of the valve in nuclear power plan. The functional test is the real test procedure and defined in KEPIC MF

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

EPRI flow-loop/in situ test program for motor-operated valves

International Nuclear Information System (INIS)

Hosler, J.F.; Dorfman, L.S.

1994-01-01

The Electric Power Research Institute is undertaking a comprehensive research program to develop and validate methods for predicting the performance of common motor-operated gate, global, and butterfly valves. To assess motor-operated valve (MOV) performance characteristics and provide a basis for methods validation, full-scale testing was conducted on 62 MOVs. Tests were performed in four flow-loop facilities and in nine nuclear units. Forty-seven gate, five globe, and 10 butterfly valves were tested under a wide range of flow and differential pressure conditions. The paper describes the test program scope, test configurations, instrumentation and data acquisition, testing approach, and data analysis methods. Key results are summarized

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)

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

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.)

Timing tests: automatic valve closure for tritium leaks

International Nuclear Information System (INIS)

Hanel, S.

1976-01-01

How fast can an automotive valve be closed after a tritium leak occurs in a system. Tests described found that a valve can be closed within fifteen seconds of leakage. In one practical example considered, this delay would limit loss of tritium from a plumbing leak in a tritium system to 1 1 / 4 g. The tests were made in a typical LLL air-flush hood in which a tritium handling system had been installed. Incidental observations suggest that further study be made of a possible leak-actuated recovery system for an entire tritium facility

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...

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

Proceedings of the symposium on inservice testing of pumps and valves

International Nuclear Information System (INIS)

1990-10-01

The 1990 Symposium on Inservice Testing of Pumps and Valves, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provided a forum for the discussion of current programs and methods for inservice testing at nuclear power plants. The symposium also provided an opportunity to discuss the need to improve inservice testing in order to ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants resulted in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants

Proceedings of the symposium on inservice testing of pumps and valves

Energy Technology Data Exchange (ETDEWEB)

1990-10-01

The 1990 Symposium on Inservice Testing of Pumps and Valves, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provided a forum for the discussion of current programs and methods for inservice testing at nuclear power plants. The symposium also provided an opportunity to discuss the need to improve inservice testing in order to ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants resulted in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants.

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

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

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)

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)

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)

Motor operated valve testing and the 'rate of loading' phenomenon

International Nuclear Information System (INIS)

Black, B.R.

1991-01-01

This paper discusses valve design features which affect the ability to predict motor operated valve (MOV) performance and reviews factors which should be considered when selecting switch settings to limit stem loads. Considerable attention is given to the rate of loading phenomenon which affects the relationship between valve stem thrust and actuator spring pack deflection. Equations are developed, and testing is discussed which permit the construction of an MOV dynamic model. Factors which must be considered when maintaining switch settings correct throughout the life of the plant are discussed. And switch setting acceptance criteria for use with baseline Static and Design Basis testing are suggested

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...

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.

A novel polymeric prosthetic heart valve: design, manufacture, and testing

OpenAIRE

Brubert, Jacob

2016-01-01

In this thesis a flexible leaflet polymeric prosthetic aortic heart valve was designed, manufactured, and tested. The prosthesis was designed with the aim of overcoming the need for anticoagulant therapy, which is required for current mechanical prostheses; while also having lifelong durability, which current bioprosthetic heart valves are not able to achieve. Inspired by the anisotropic architecture of collagen in the natural valve, a shortlist of polystyrene based block copolymers (BCPs), w...

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

RETRAN simulation of Oyster Creek MSIV closure and bypass valve tests

International Nuclear Information System (INIS)

Alammar, M.A.

1987-01-01

A series of benchmarks against start-up tests have been performed on the Oyster Creek boiling water reactor unit 2 RETRAN model in support of developing an in-house reload capability. The liquid and the pressure regulator models have been benchmarked against level and pressure setpoint changes, where small setpoint perturbations were made at rated power. The purpose of the present benchmark is to check the liquid level behavior during a severe level drop as during void collapse following a scram and to size the bypass valves by benchmarking the valves' contraction coefficient. The main steam isolation valves (MSIVs) closure start-up test was chosen for the former, while the bypass valve test was chosen for the latter. The two benchmarks complete the qualification of the upper downcomer liquid level for small and large level changes and the pressure regulator system for the Oyster Creek RETRAN model

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...

Proceedings of the 4th NRC/ASME symposium on valve and pump testing

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-08-01

The 1996 Symposium on Valve and Pump Testing, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the U.S. Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. Individual papers of this Proceedings have been cataloged separately.

Proceedings of the 4th NRC/ASME symposium on valve and pump testing

International Nuclear Information System (INIS)

1996-01-01

The 1996 Symposium on Valve and Pump Testing, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the U.S. Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. Individual papers of this Proceedings have been cataloged separately

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

Implementation of an enlarged model of the safety valves and relief in the plant integral model for the code RELAP/SCDAPSIM

International Nuclear Information System (INIS)

Amador G, R.; Ortiz V, J.; Castillo D, R.; Hernandez L, E. J.; Galeana R, J. C.; Gutierrez, V. H.

2013-10-01

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)

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

International Nuclear Information System (INIS)

Fuller, R.; Harrell, J.

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Fuller, R.; Harrell, J.

1996-12-01

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

Major results from safety-related integral effect tests with VISTA-ITL for the SMART design

International Nuclear Information System (INIS)

Park, H. S.; Min, B. Y.; Shin, Y. C.; Yi, S. J.

2012-01-01

A series of integral effect tests (IETs) was performed by the Korea Atomic Energy Research Inst. (KAERI) using the VISTA integral test loop (VISTA-ITL) as a small-scale IET program. Among them this paper presents major results acquired from the safety-related IETs with the VISTA-ITL facility for the SMART design. Three small-break loss-of-coolant accident (SBLOCA) tests of safety injection system (SIS) line break, shutdown cooling system (SCS) line break and pressurizer safety valve (PSV) line break were successfully performed and the transient characteristics of a complete loss of flowrate (CLOF) was simulated properly with the VISTA-ITL facility. (authors)

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

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...

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.

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

Valve leakage inspection testing and maintenance process

International Nuclear Information System (INIS)

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

1991-01-01

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

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...

Preliminary design for hot dirty-gas control-valve test facility. Final report

Energy Technology Data Exchange (ETDEWEB)

1980-01-01

This report presents the results of a preliminary design and cost estimating effort for a facility for the testing of control valves in Hot Dirty Gas (HDGCV) service. This design was performed by Mittelhauser Corporation for the United States Department of Energy's Morgantown Energy Technology Center (METC). The objective of this effort was to provide METC with a feasible preliminary design for a test facility which could be used to evaluate valve designs under simulated service conditions and provide a technology data base for DOE and industry. In addition to the actual preliminary design of the test facility, final design/construction/operating schedules and a facility cost estimate were prepared to provide METC sufficient information with which to evaluate this design. The bases, assumptions, and limitations of this study effort are given. The tasks carried out were as follows: METC Facility Review, Environmental Control Study, Gas Generation Study, Metallurgy Review, Safety Review, Facility Process Design, Facility Conceptual Layout, Instrumentation Design, Cost Estimates, and Schedules. The report provides information regarding the methods of approach used in the various tasks involved in the completion of this study. Section 5.0 of this report presents the results of the study effort. The results obtained from the above-defined tasks are described briefly. The turnkey cost of the test facility is estimated to be $9,774,700 in fourth quarter 1979 dollars, and the annual operating cost is estimated to be $960,000 plus utilities costs which are not included because unit costs per utility were not available from METC.

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

Cavitation behavior observed in three monoleaflet mechanical heart valves under accelerated testing conditions.

Science.gov (United States)

Lo, Chi-Wen; Liu, Jia-Shing; Li, Chi-Pei; Lu, Po-Chien; Hwang, Ned H

2008-01-01

Accelerated testing provides a substantial amount of data on mechanical heart valve durability in a short period of time, but such conditions may not accurately reflect in vivo performance. Cavitation, which occurs during mechanical heart valve closure when local flow field pressure decreases below vapor pressure, is thought to play a role in valve damage under accelerated conditions. The underlying flow dynamics and mechanisms behind cavitation bubble formation are poorly understood. Under physiologic conditions, random perivalvular cavitation is difficult to capture. We applied accelerated testing at a pulse rate of 600 bpm and transvalvular pressure of 120 mm Hg, with synchronized videographs and high-frequency pressure measurements, to study cavitation of the Medtronic Hall Standard (MHS), Medtronic Hall D-16 (MHD), and Omni Carbon (OC) valves. Results showed cavitation bubbles between 340 and 360 micros after leaflet/housing impact of the MHS, MHD, and OC valves, intensified by significant leaflet rebound. Squeeze flow, Venturi, and water hammer effects each contributed to cavitation, depending on valve design.

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.

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

Effects on fatigue life of gate valves due to higher torque switch settings during operability testing

International Nuclear Information System (INIS)

Richins, W.D.; Snow, S.D.; Miller, G.K.; Russell, M.J.; Ware, A.G.

1995-12-01

Some motor operated valves now have higher torque switch settings due to regulatory requirements to ensure valve operability with appropriate margins at design basis conditions. Verifying operability with these settings imposes higher stem loads during periodic inservice testing. These higher test loads increase stresses in the various valve internal parts which may in turn increase the fatigue usage factors. This increased fatigue is judged to be a concern primarily in the valve disks, seats, yokes, stems, and stem nuts. Although the motor operators may also have significantly increased loading, they are being evaluated by the manufacturers and are beyond the scope of this study. Two gate valves representative of both relatively weak and strong valves commonly used in commercial nuclear applications were selected for fatigue analyses. Detailed dimensional and test data were available for both valves from previous studies at the Idaho National Engineering Laboratory. Finite element models were developed to estimate maximum stresses in the internal parts of the valves and to identity the critical areas within the valves where fatigue may be a concern. Loads were estimated using industry standard equations for calculating torque switch settings prior and subsequent to the testing requirements of USNRC Generic Letter 89--10. Test data were used to determine both; (1) the overshoot load between torque switch trip and final seating of the disk during valve closing and (2) the stem thrust required to open the valves. The ranges of peak stresses thus determined were then used to estimate the increase in the fatigue usage factors due to the higher stem thrust loads. The usages that would be accumulated by 100 base cycles plus one or eight test cycles per year over 40 and 60 years of operation were calculated

NRC test results and operations experience provide insights for a new gate valve stem force correlation

International Nuclear Information System (INIS)

Watkins, John C.; Steele, Robert Jr.; DeWall, Kevin G.; Weidenhamer, G.H.; Rothberg, O.O.

1994-01-01

This paper presents the results of testing sponsored by the NRC to assess valve and motor operator performance under varying pressure and fluid conditions. This effort included an examination of the methods used by the industry to predict the required stem force of a valve, and research to provide guidelines for the extrapolation of in situ test results to design basis conditions.Years ago, when most of these valves were originally installed, the industry used a set of equations to determine analytically that the valves' motor-operators were large enough and the control switches were set high enough to close the valves at their design basis conditions. Our research has identified several inconsistencies with the industry's existing gate valve stem force equation and has challenged the overly simplistic assumptions inherent in its use. This paper discusses the development of the INEL correlation, which serves as the basis for a method to bound the stem force necessary to close flexwedge gate valves whose operational characteristics have been shown to be predictable. As utilities undertake to provide assurance of their valves' operability, this ability to predict analytically the required stem force is especially important for valves that cannot be tested at design basis conditions. For such valves, the results of tests conducted at less severe conditions can be used with the INEL correlation to make the necessary prediction. ((orig.))

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)

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.)

Use of expert judgment in the development and evaluation of risk-based inservice testing strategies for pumps and valves

International Nuclear Information System (INIS)

McAllister, W.J.; Perdue, R.K.; Balkey, K.R.; Closky, N.B.

1996-01-01

This paper describes a rigorous approach for quantitatively evaluating inservice testing effectiveness that evolved from two pilot plant studies. These studies prototyped methodologies for designing and selecting inservice testing (IST) strategies in a manner structured to insure that the targeted components will perform their required safety functions while minimizing life cycle inservice testing costs. The paper concentrates on the use of expert judgment in developing test effectiveness measures that move risk-based methods beyond ranking to optimization of plant IST programs. Selected results for check valves and pumps are shown to illustrate the practical significance of the approach

Use of expert judgment in the development and evaluation of risk-based inservice testing strategies for pumps and valves

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

This paper describes a rigorous approach for quantitatively evaluating inservice testing effectiveness that evolved from two pilot plant studies. These studies prototyped methodologies for designing and selecting inservice testing (IST) strategies in a manner structured to insure that the targeted components will perform their required safety functions while minimizing life cycle inservice testing costs. The paper concentrates on the use of expert judgment in developing test effectiveness measures that move risk-based methods beyond ranking to optimization of plant IST programs. Selected results for check valves and pumps are shown to illustrate the practical significance of the approach.

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

Fast flux test facility primary sodium check valve

International Nuclear Information System (INIS)

Rabe, G.B.; Nash, C.F.

1975-01-01

The design and development of a tilting-disc check valve for the primary sodium coolant loop of the Fast Flux Test Facility is described. The demanding design requirements specified for this system dictated a design with unique features. These features, along with the structural design and analysis requirements and the testing program used to develop and justify the design, are described

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

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.

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

Acoustic-Modal Testing of the Ares I Launch Abort System Attitude Control Motor Valve

Science.gov (United States)

Davis, R. Benjamin; Fischbach, Sean R.

2010-01-01

The Attitude Control Motor (ACM) is being developed for use in the Launch Abort System (LAS) of NASA's Ares I launch vehicle. The ACM consists of a small solid rocket motor and eight actuated pintle valves that directionally allocate.thrust_- 1t.has-been- predicted-that significant unsteady. pressure.fluctuations.will.exist. inside the-valves during operation. The dominant frequencies of these oscillations correspond to the lowest several acoustic natural frequencies of the individual valves. An acoustic finite element model of the fluid volume inside the valve has been critical to the prediction of these frequencies and their associated mode shapes. This work describes an effort to experimentally validate the acoustic finite model of the valve with an acoustic modal test. The modal test involved instrumenting a flight-like valve with six microphones and then exciting the enclosed air with a loudspeaker. The loudspeaker was configured to deliver broadband noise at relatively high sound pressure levels. The aquired microphone signals were post-processed and compared to results generated from the acoustic finite element model. Initial comparisons between the test data and the model results revealed that additional model refinement was necessary. Specifically, the model was updated to implement a complex impedance boundary condition at the entrance to the valve supply tube. This boundary condition models the frequency-dependent impedance that an acoustic wave will encounter as it reaches the end of the supply tube. Upon invoking this boundary condition, significantly improved agreement between the test data and the model was realized.

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

Functional and performance evaluation of 28 bar hot shutdown passive valve (HSPV) at integral test loop (ITL) for advanced heavy water reactor (AHWR)

International Nuclear Information System (INIS)

Sapra, M.K.; Kundu, S.; Pal, A.K.; Sharma, B.S.V.G.

2007-02-01

During reactor shutdown in advanced heavy water reactor (AHWR), core decay heat is removed by eight isolation condensers (IC) submerged in gravity driven water pool. Passive valves are provided on the down stream of each isolation condenser. On increase in steam drum pressure beyond a set value, these passive valves start opening and establish steam flow by natural circulation between the four steam drums and corresponding isolation condensers under hot shutdown and therefore they are termed as Hot Shut Down Passive Valves (HSPVs). The HSPV is a self acting type valve requiring no external energy, i.e. neither air nor electric supply for actuation. This feature makes the valve functioning independent of external systems such as compressed air supply or electric power supply, thereby providing inherent safety feature in line with reactor design philosophy. The high pressure and high temperature HSPV s for nuclear reactor use, are non-standard valves and therefore not manufactured by the valve industry worldwide. In the process of design and development of a prototype valve for AHWR, a 28 bar HSPV was configured and successfully tested at Integral Test Loop (ITL) at Engineering Hall No.7. During ten continuous experiments spread over 14 days, the HSPV has proved its functional capabilities and its intended use in decay heat removal system. The in-situ pressure setting and calibration aspect of HSPV has also been successfully established during these experiments. This report gives an insight into the HSPV's functional behavior and role in reactor decay heat removal system. The report not only provides the quantitative measure of performance for 28 bar HSPV in terms of valve characteristics, pressure controllability, linearity and hysteresis but also sets qualitative indicators for prototype 80 bar HSPV, being developed for AHWR. (author)

Test to prove the resistance to incidents of components of electric and control systems in the safety containment of nuclear power plants

International Nuclear Information System (INIS)

1982-01-01

The marginal program for proving the suitability of safety-relevant components of electric and control systems in the safety containment during a loss-of-coolant incident is described. Variant test conditions are established in the component-specific test program. Special attention has been paid to the representation of the course of pressure and temperature for the performance test of the valve room of the Nuclear Power Plant Philippsburg 2. (DG) [de

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

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)

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.

Textile for heart valve prostheses: fabric long-term durability testing.

Science.gov (United States)

Heim, Frederic; Durand, Bernard; Chakfe, Nabil

2010-01-01

The rapid developments and success in percutaneous vascular surgery over the last two decades with the now common stent grafts implantation, make the noninvasive surgery technique today attractive even for heart valve replacement. Less traumatic for the patient and also less time consuming, percutaneous heart valve replacement is however at its beginning and restricted to end of life patients. The noninvasive procedure expects from the heart valve prosthesis material to be resistant and adapted to folding requirements of the implantation process (catheter). Polyester fabric could be a suited material for heart valve implanted percutaneously. Highly flexible and resistant, polyester fabric proved to be well adapted to the dynamic behavior of a valve and polyester (Dacron) is also widely used for vascular grafts implantation and shows good biocompatibility and durability. However, today there's no data available on long-term durability of fabric used as heart valve material. The purpose of this work is to study the long term behavior of a microdenier polyester fabric construction under combined in vitro flexure and tension fatigue stress. In the novel in vitro testing technique presented, a fabric specimen was subjected to combined flexural and tensile fatigue generated by fluid flow under physiological pressure conditions. The results obtained show how flexural properties change with fatigue time, which reflects directly on the suitability of a fabric in such devices. It was also observed that these fabric structural changes directly influence the in vitro behavior of the textile heart valve prosthesis. (c) 2009 Wiley Periodicals, Inc.

A novel synthetic test system for thyristor level in the converter valve of HVDC power transmission

Directory of Open Access Journals (Sweden)

Liu Longchen

2016-01-01

Full Text Available The converter valve is the core equipment in the HVDC power transmission system, a+-nd its performance has a direct effect on the reliability, stability and efficiency of the whole power system. As the basic unit of HVDC converter valve, the thyristor level needs to be test routinely in order to grasp the state of the converter valve equipment. Therefore, it is urgent to develop a novel synthetic test system for the thyristor level with thyristor control unit (TCU. However, currently there is no specific test scheme for the thyristor level of HVDC converter valve. In this paper, the synthetic test principle, content and methods for the thyristor level with TCU are presented based on the analysis of the thyristor reverse recovery characteristic and the IEC technology standard. And a transient high-voltage pulse is applied to the thyristor level during its reverse recovery period in order to test the characteristics of thyristor level. Then, the synthetic test system for the thyristor level is applied to the converter valve test of ±800 kV HVDC power transmission project, and the practical test result verifies the reasonability and validity of the proposed synthetic test system.

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

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

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

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

Aggressive chemical decontamination tests on small valves from the Garigliano BWR

International Nuclear Information System (INIS)

Bregani, F.

1990-01-01

In order to check the effectiveness of direct chemical decontamination on small and complex components, usually considered for storage without decontamination because of the small amount, some tests were performed on the DECO experimental loop. Four small stainless steel valves from the primary system of the Garigliano BWR were decontaminated using mainly aggressive chemicals such as HC1, HF, HNO 3 and their mixtures. On two valves, before the treatment with aggressive chemicals, a step with soft chemical (oxalic and citric acid mixture) was performed in order to see whether a softening action enhances the following aggressive decontamination. Moreover, in order to increase as much as possible the decontamination effectiveness, a decontamination process using ultrasounds jointly with aggressive chemicals was investigated. After an intensive laboratory testing programme, two smaller stainless steel valves from the primary system of the Garigliano BWR were decontaminated using ultrasounds in aggressive chemical solutions

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

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

OpenAIRE

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

2017-01-01

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

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

CCF analysis of high redundancy systems safety/relief valve data analysis and reference BWR application

International Nuclear Information System (INIS)

Mankamo, T.; Bjoere, S.; Olsson, Lena

1992-12-01

Dependent failure analysis and modeling were developed for high redundancy systems. The study included a comprehensive data analysis of safety and relief valves at the Finnish and Swedish BWR plants, resulting in improved understanding of Common Cause Failure mechanisms in these components. The reference application on the Forsmark 1/2 reactor relief system, constituting of twelve safety/relief lines and two regulating relief lines, covered different safety criteria cases of reactor depressurization and overpressure protection function, and failure to re close sequences. For the quantification of dependencies, the Alpha Factor Model, the Binomial Probability Model and the Common Load Model were compared for applicability in high redundancy systems

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

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

Operating reliability of valves in French pressurized water nuclear power plants

International Nuclear Information System (INIS)

Conte

1986-10-01

Taking into account the large numbers of valves (about 10000) of a PWR nuclear power plant, the importance of some valves in the safety functions and the cost resulting from their unavailability, the individual operability of these equipments has to be ensured at a high reliability level. This assurance can be obtained by means of an effort at all the stages which contribute to the quality of the product: design, qualification tests, fabrication, tests at the start-up stage, maintenance and tests during the power plant operation, experience feedback. This paper emphasizes more particularly on the tests carried out on loops of qualification [fr

Project W-314 acceptance test report HNF-4643 for HNF-4642 241-AN-A valve pit manifold valves and position indication for project W-314

International Nuclear Information System (INIS)

HAMMERS, J.S.

1999-01-01

The purpose of the test was to verify that the AN Tank Farm Manifold Valves can be manually manipulated to the required operating position and that the electrical and visual indications accurately reflect that position. Physical locking devices were also verified to function. The Acceptance Test Procedure HNF-4642, 241-AN-A Valve Pit Manifold Valves and Position Indication was conducted between 23 June and 10 August 1999 at the 200E AN Tank Farm. The test has no open test exceptions. The test was conducted prior to final engineering ''as built'' activities being completed, this had an impact on the procedure and test results, ECN 653752 was written to correct the mismatch between the procedure and actual field conditions. P and ID H-14-100941 was changed via ECN-W-314-4C-120. All components, identified in the procedure, were not found to be labeled and identified as written in the procedure, temporary tags were used for operational identification. A retest of valve ANA-WT-V 318 was required because it was removed from its installed position and modified after testing was completed

Comparative study of Butterfly valves

International Nuclear Information System (INIS)

Galmes Belmonte, F.B.

1998-01-01

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

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

Safety relief valve instabilities; Instabiles Verhalten beim Betrieb von direkt belasteten Sicherheitsventilen

Energy Technology Data Exchange (ETDEWEB)

Neuhaus, Thorsten; Bloemeling, Frank; Jung, Andreas; Schaffrath, Andreas [TUeV NORD SysTec GmbH und Co. KG, Hamburg (Germany)

2012-11-01

The opening and closing of safety relief valves induce pressure vibrations and respective loads on the piping system. The quantification of the internal pressure load and the pipe segment forces for the proof of the structural integrity of the piping system and the supports is performed using the code DYVRO. It is not clear whether the calculated result of high-frequent opening and closing shows a realistic behavior or is caused by a simplified modeling. The contribution offers strategic recommendations to avoid unrealistic calculations.

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)

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.)

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

International Nuclear Information System (INIS)

Wheeler, A.J.

1978-02-01

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

Leak test method for radioactive material packagings without pressure valve connections

International Nuclear Information System (INIS)

Johnson, S.F.; Stenbaeck, A.

1976-01-01

A leak test method has been developed at Studsvik which provides the possibility of testing Type B packagings unequipped with valves for evacuation or pressurizing. Even large packagings with pressure valve connections can be leak tested by this method. The method is a pressure test method. The test gas comprises a mixture of helium and nitrogen or helium and air. Excess pressure in a valveless packaging is achieved by vaporization of liquid nitrogen. All parts of the packaging or package where leaks might be expected are covered by plastic sheet. Samples of the gas accumulated under the plastic sheets are taken using evacuated glass ampoules which are initially sealed off to a breakable point. The gas samples are measured with a He-mass spectrometer. The sensitivity of this method of leak testing is, in practice, of the order of 10 -7 atmcm 3 s -1 . (author)

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

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

A Methodology for Evaluation of Inservice Test Intervals for Pumps and Motor-Operated Valves

International Nuclear Information System (INIS)

Cox, D.F.; Haynes, H.D.; McElhaney, K.L.; Otaduy, P.J.; Staunton, R.H.; Vesely, W.E.

1999-01-01

Recent nuclear industry reevaluation of component inservice testing (IST) requirements is resulting in requests for IST interval extensions and changes to traditional IST programs. To evaluate these requests, long-term component performance and the methods for mitigating degradation need to be understood. Determining the appropriate IST intervals, along with component testing, monitoring, trending, and maintenance effects, has become necessary. This study provides guidelines to support the evaluation of IST intervals for pumps and motor-operated valves (MOVs). It presents specific engineering information pertinent to the performance and monitoring/testing of pumps and MOVs, provides an analytical methodology for assessing the bounding effects of aging on component margin behavior, and identifies basic elements of an overall program to help ensure component operability. Guidance for assessing probabilistic methods and the risk importance and safety consequences of the performance of pumps and MOVs has not been specifically included within the scope of this report, but these elements may be included in licensee change requests

Project W-314 specific test and evaluation plan for 241-AN-A valve pit

International Nuclear Information System (INIS)

Hays, W.H.

1997-01-01

The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AN-A Valve Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a ''lower tier'' document based on the W-314 Test and Evaluation Plan (TEP) This STEP encompasses all testing activities required to demonstrate compliance to the project design criteria as it relates to the modifications of the AN-A valve pit. The Project Design Specifications (PDS) identify the specific testing activities required for the Project. Testing includes Validations and Verifications (e.g., Commercial Grade Item Dedication activities), Factory Acceptance Tests (FATs), installation tests and inspections, Construction Acceptance Tests (CATs), Acceptance Test Procedures (ATPs), Pre-Operational Test Procedures (POTPs), and Operational Test Procedures (OTPs). It should be noted that POTPs are not required for testing of the modifications to the 241-AN-A Valve Pit. The STEP will be utilized in conjunction with the TEP for verification and validation

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...

Development of a system for monitoring and diagnosis using Fuzzy logic in control valves of laboratory test equipment of Experimental Center Aramar

International Nuclear Information System (INIS)

Porto Junior, Almir Carlos Soares

2014-01-01

The question of components reliability, specifically process control valves, has become an important issue to be investigated in nuclear power plants and other areas such as refinery or offshore oil rig, considering the safety and life extension of the plant. The development of non intrusive monitoring and diagnostic method allows the identification of defects in components of the plant during normal operation. The objective of this dissertation is to present an analysis and diagnosis of control valves of a steam plant part that simulates the secondary circuit of a pressurized water reactor. This installation is part of propulsion equipment testing laboratory of the Brazilian Navy, at Ipero-SP. The methodology for design is based on graphical analysis of two parameters, the valve air pressure actuator and the displacement of the valve plug. These data are extracted by a smart positioner, part of Delta V™ Automation System. An analysis is implemented in detecting anomalies by an approach using Expert Systems by the technique of fuzzy logic. Once the basic measures of control valves are taken, it is possible to detect symptoms of failure, leakage, friction, damage, etc. The monitoring and diagnostic system has been designed in MATLAB® version 2009 th by the complement 'Fuzzy Logic Toolbox'. It is a noninvasive technique. Thus, it is possible to know what is happening with the chosen components, just analyzing the parameters of the valve. The software called ValveLink® (developed by Emerson) receives signals from hardware component (intelligent positioner) installed next to the control valve. These signals (electrical current) are transformed into information which are used input parameters: air pressure valve actuator and valve plug displacement. With the use of fuzzy logic, these parameters are interpreted. They suffer inferences by rules written by experts in valves. After these inferences, the information is processed and sent as output signals

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

The Performance Test for Reactor Coolant Pump (RCP) adopting Variable Restriction Orifice Type Control Valve

Energy Technology Data Exchange (ETDEWEB)

Kim, S.; Bae, B. U.; Cho, Y. J. and others

2014-05-15

The design values of the RCPTF are 17.2 MPa, 343 .deg. C, 11.7 m{sup 3}/s, and 13 MW in the maximum pressure, temperature, flow rate, and electrical power, respectively. In the RCPTF, various types of tests can be performed including a hydraulic performance test to acquire a H-Q curve as well seal transient tests, thrust bearing transient test, cost down test, NPSHR verification test, and so on. After a commissioning startup test was successfully perfomed, mechanical structures are improved including a flow stabilizer and variable restriction orifice. Two- branch pipe (Y-branch) was installed to regulate the flow rate in the range of performance tests. In the main pipe, a flow restrictor (RO: Restriction Orifice) for limiting the maximum flow rate was installed. In the branch pipe line, a globe valve and a butterfly valves for regulating the flow rate was located on the each branch line. When the pressure loss of the valve side is smaller than that of the RO side, the flow rate of valve side was increasing and the flow disturbance was occurred in the lower pipe line. Due to flow disturbnace, it is to cause an error when measuring RCP head and flow measurement of the venturi flow meter installed in the lower main pipe line, and thus leading to a decrease in measurement accuracy as a result. To increase the efficiency of the flow control availability of the test facility, the variable restriction orifice (VRO) type flow control valve was designed and manufactured. In the RCPTF in KAERI, the performance tests and various kinds of transient tests of the RCP were successfully performed. In this study, H-Q curve of the pump using the VRO revealed a similar trend to the result from two ROs. The VRO was confirmed to effectively cover the full test range of the flow rate.

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

Gate valve performance prediction

International Nuclear Information System (INIS)

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

1994-01-01

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

Valve for the mechanical isolation of a pipe to take up a test probe

International Nuclear Information System (INIS)

Uecker, D.F.

1976-01-01

A valve is introduced for application in a pipe in which a test probe is arranged. The valve serves to isolate the pipe in a gas-tight way, thus preventing the escape of radioactive gas or dust during operation in a nuclear reactor. (TK) [de

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)

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

NRC inspections of licensee activities to improve the performance of motor-operated valves

International Nuclear Information System (INIS)

Scarbrough, T.G.

1992-01-01

The NRC regulations require that components important to the safe operation of a nuclear power plant be treated in a manner that provides assurance of their proper performance. Despite these regulatory requirements, operating experience and research programs have raised concerns regarding the performance of motor-operated valves (MOVs) in nuclear power plants. In June 1990, the staff issued NUREG-1352, Action Plans for Motor-Operated Valves and Check Valves, which contains planned actions to organize the activities aimed at resolving the concerns about MOV performance. A significant task of the MOV action plan is the staff's review of the implementation of Generic Letter (GL) 89-10 (June 28, 1989), 'Safety-Related Motor-Operated Valve Testing and Surveillance,' and its supplements, by nuclear power plant licensees. The NRC staff has issued several supplements to GL 89-10 to provide additional guidance for use by licensees in responding to the generic letter. The NRC staff has conducted initial inspections of the GL 89-10 programs at most licensee facilities. This paper outlines some of the more significant findings of those inspections. For example, licensees who have begun differential pressure and flow testing have found some MOVs to require more thrust to operate than predicted by the standard industry equation with typical valve factors assumed in the past. The NRC staff has found weaknesses in licensee procedures for conducting the differential pressure and flow tests, the acceptance criteria for the tests in evaluating the capability of the MOV to perform its safety function under design basis conditions, and feedback of the test results into the methodology used by the licensee in predicting the thrust requirements for other MOVs. Some licensees have not made adequate progress toward resolving the MOV issue for their facilities within the recommended schedule of GL 89-10

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

Motor-operated valve (MOV) actuator motor and gearbox testing

International Nuclear Information System (INIS)

DeWall, K.; Watkins, J.C.; Bramwell, D.

1997-07-01

Researchers at the Idaho National Engineering and Environmental Laboratory tested the performance of electric motors and actuator gearboxes typical of the equipment installed on motor-operated valves used in nuclear power plants. Using a test stand that simulates valve closure loads against flow and pressure, the authors tested five electric motors (four ac and one dc) and three gearboxes at conditions a motor might experience in a power plant, including such off-normal conditions as operation at high temperature and reduced voltage. They also monitored the efficiency of the actuator gearbox. All five motors operated at or above their rated starting torque during tests at normal voltages and temperatures. For all five motors, actual torque losses due to voltage degradation were greater than the losses calculated by methods typically used for predicting motor torque at degraded voltage conditions. For the dc motor the actual torque losses due to elevated operating temperatures were greater than the losses calculated by the typical predictive method. The actual efficiencies of the actuator gearboxes were generally lower than the running efficiencies published by the manufacturer and were generally nearer the published pull-out efficiencies. Operation of the gearbox at elevated temperature did not affect the operating efficiency

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

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Testing of valves and associated systems in large scale experiments

International Nuclear Information System (INIS)

Becker, M.

1985-01-01

The system examples dealt with are selected so that they cover a wide spectrum of technical tasks and limits. Therefore the flowing medium varies from pure steam flow via a mixed flow of steam and water to pure water flow. The valves concerned include those whose main function is opening, and also those whose main function is the secure closing. There is a certain limitation in that the examples are taken from Boiling Water Reactor technology. The main procedure in valve and system testing described is, of course, not limited to the selected examples, but applies generally in powerstation and process technology. (orig./HAG) [de

Test quality

International Nuclear Information System (INIS)

Hartley, R.S.; Keller, A.E.

1992-01-01

Inservice testing of safety-related components at nuclear power plants is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section 11 of the Code state test method and frequency requirements for pumps and valves, respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test's ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing

Statistical characteristics of mechanical heart valve cavitation in accelerated testing.

Science.gov (United States)

Wu, Changfu; Hwang, Ned H C; Lin, Yu-Kweng M

2004-07-01

Cavitation damage has been observed on mechanical heart valves (MHVs) undergoing accelerated testing. Cavitation itself can be modeled as a stochastic process, as it varies from beat to beat of the testing machine. This in-vitro study was undertaken to investigate the statistical characteristics of MHV cavitation. A 25-mm St. Jude Medical bileaflet MHV (SJM 25) was tested in an accelerated tester at various pulse rates, ranging from 300 to 1,000 bpm, with stepwise increments of 100 bpm. A miniature pressure transducer was placed near a leaflet tip on the inflow side of the valve, to monitor regional transient pressure fluctuations at instants of valve closure. The pressure trace associated with each beat was passed through a 70 kHz high-pass digital filter to extract the high-frequency oscillation (HFO) components resulting from the collapse of cavitation bubbles. Three intensity-related measures were calculated for each HFO burst: its time span; its local root-mean-square (LRMS) value; and the area enveloped by the absolute value of the HFO pressure trace and the time axis, referred to as cavitation impulse. These were treated as stochastic processes, of which the first-order probability density functions (PDFs) were estimated for each test rate. Both the LRMS value and cavitation impulse were log-normal distributed, and the time span was normal distributed. These distribution laws were consistent at different test rates. The present investigation was directed at understanding MHV cavitation as a stochastic process. The results provide a basis for establishing further the statistical relationship between cavitation intensity and time-evolving cavitation damage on MHV surfaces. These data are required to assess and compare the performance of MHVs of different designs.

Leak testing in isolation in Almaraz NPP valves; Pruebas de fugas en valvulas de aislamiento en C.N. Almaraz

Energy Technology Data Exchange (ETDEWEB)

Garcia Rocasolano, J.; Delgado Flores, E.

2014-07-01

The components, object of this paper are isolation valves of the containment of waste heat, injection safety and water extraction systems of sealing of the main pumps. A review and improvement of the methodology applied by means of an analysis of the provision and operation of certain systems, as well as practice followed by other plants in the US, aims to achieve a more effective testing program by reducing doses to staff, the generation of radioactive waste by drainage systems, and the risks of introduction of parts loose in the primary. (Author)

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

On-line valve monitoring at the Ormen Lange gas plant

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

On-line valve monitoring at the Ormen Lange gas plant

International Nuclear Information System (INIS)

Greenlees, R.; Hale, S.

2011-01-01

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

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.

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.

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

Small sodium valve design and operating experience

International Nuclear Information System (INIS)

McGough, C.B.

1974-01-01

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

Testing of one-inch UF{sub 6} cylinder valves under simulated fire conditions

Energy Technology Data Exchange (ETDEWEB)

Elliott, P.G. [Martin Marietta Energy Systems, Inc., Paducah, KY (United States)

1991-12-31

Accurate computational models which predict the behavior of UF{sub 6} cylinders exposed to fires are required to validate existing firefighting and emergency response procedures. Since the cylinder valve is a factor in the containment provided by the UF{sub 6} cylinder, its behavior under fire conditions has been a necessary assumption in the development of such models. Consequently, test data is needed to substantiate these assumptions. Several studies cited in this document provide data related to the behavior of a 1-inch UF{sub 6} cylinder valve in fire situations. To acquire additional data, a series of tests were conducted at the Paducah Gaseous Diffusion Plant (PGDP) under a unique set of test conditions. This document describes this testing and the resulting data.

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)

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

International Nuclear Information System (INIS)

Haynes, H.D.

1988-01-01

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

Characteristic analysis of servo valve

International Nuclear Information System (INIS)

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

2008-01-01

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

Application of ceramics to the sliding seat of valve bridge; Valve bridge yodobu eno ceramics tekiyo

Energy Technology Data Exchange (ETDEWEB)

Matsui, T; Ono, T [Mitsubishi Motors Corp., Tokyo (Japan)

1997-10-01

For use in the valve train, using an OHV (over head valve) configuration. of a 4 valve diesel engine for trucks and buses; we developed a valve bridge, a component of a valve train, with a ceramic head that is made of silicon nitride(Si3N4) in contact with a rocker arm in order to reduce cost and improve wear resistance for further diesel engine emissions regulations. In order to evaluate the effect of this valve bridge, RIG tests and durability tests on actual engines were carried out. 7 figs., 2 tabs.

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

Test quality

International Nuclear Information System (INIS)

Hartley, R.S.; Keller, A.E.

1992-01-01

This document discusses inservice testing of safety-related components at nuclear power plants which is performed under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (the Code). Subsections IWP and IWV of Section XI of the Code state test method and frequency requirements for pumps and valves respectively. Tests vary greatly in quality and frequency. This paper explores the concept of test quality and its relationship with operational readiness and preventive maintenance. This paper also considers the frequencies of component testing. Test quality is related to a test's ability to detect degradation that can cause component failure. The quality of the test depends on several factors, including specific parameters measured, system or component conditions, and instrument accuracy. The quality of some currently required tests for check valves, motor-operated valves, and pumps is also discussed. Suggestions are made to improve test quality by measuring different parameters, testing valves under load, and testing positive displacement pumps at high pressure and centrifugal pumps at high flow rate conditions. These suggestions can help to improve the level of assurance of component operational readiness gained from testing

Thermal tests on UF6 containers and valves modelisation and extrapolation on real fire situations

International Nuclear Information System (INIS)

Duret, B.; Warniez, P.

1988-12-01

From realistic tests on containers or on valves, we propose a modelisation which we apply to 3 particular problems: resistance of a 48 Y containers, during a fire situation. Influence of the presence of a valve. Evaluation of a leakage through a breach, mechanically created before a fire

Hvdc valve development outstrips testing capacity: new synthetic plant at Marchwood

Energy Technology Data Exchange (ETDEWEB)

1966-01-07

A new test plant for high voltage direct current converters is being built at the CEGB's Marchwood Engineering Laboratories. Valve development has outstripped the 60 kV 600 A bridge rating of the existing zero power factor plant.

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...

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.

The online sealing performance test of the primary circuit pressure boundary check valve in nuclear power plants

International Nuclear Information System (INIS)

Yang Yunfei; Huang Huimin

2013-01-01

The primary circuit pressure boundary check valves of 320 MW pressurized water reactor is a nuclear grade I key equipment. The sealing demand is very high, which is directly related to the internal leakage rate of the primary circuit system. After the welding check valve is repaired, the sealing performance is judged by color printing checks. If there is water or humid vapor in the pipe, it will affect the accuracy of the color printing checks. For the particularity of the online check valve tightness test, online detecting device is designed by the hydraulic pressure drop method in other nuclear power plants, but the method has some shortcomings and restrictions. In this paper, we design a reliable and portable test equipment by the low-pressure gas seal test flow measurement, which make accurate and quantitative judgment of sealing property after the pressure boundary check valves are repaired. (authors)

Butterfly valve of all rubber lining type

International Nuclear Information System (INIS)

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

1982-01-01

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

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.

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.

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)

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.)

Small valve area index: its influence on early mortality after mitral valve replacement

NARCIS (Netherlands)

Yazdanbakhsh, A. P.; van den Brink, R. B.; Dekker, Egbert; de Mol, B. A.

2000-01-01

OBJECTIVE: To test the hypothesis that mitral valve prosthesis-patient mismatch increases postoperative mortality. METHODS AND RESULTS: The effect of mitral valve prosthesis-patient mismatch on survival in a cohort of consecutive patients after mitral valve replacement with a mechanical prosthesis

Latest design of gate valves

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

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

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.)

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.

Butterfly valve torque prediction methodology

International Nuclear Information System (INIS)

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

1994-01-01

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

Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

International Nuclear Information System (INIS)

1994-07-01

The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately

Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

Energy Technology Data Exchange (ETDEWEB)

1994-07-01

The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately.

Additively Manufactured Main Fuel Valve Housing

Science.gov (United States)

Eddleman, David; Richard, Jim

2015-01-01

Selective Laser Melting (SLM) was utilized to fabricate a liquid hydrogen valve housing typical of those found in rocket engines and main propulsion systems. The SLM process allowed for a valve geometry that would be difficult, if not impossible to fabricate by traditional means. Several valve bodies were built by different SLM suppliers and assembled with valve internals. The assemblies were then tested with liquid nitrogen and operated as desired. One unit was also burst tested and sectioned for materials analysis. The design, test results, and planned testing are presented herein.

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)

Block 4 and Cluster D - Safety

CERN Multimedia

CERN. Geneva

2016-01-01

After transfer of Bloc4 test facility to SM18 its safety system was upgraded in 2014/2015 for its 3 existing test cryostats. In 2016 a new test facility HFM (High Field Magnet) was integrated in the Bloc4. A new test facility Cluster D which is similar to HFM is in a process of installation in SM18. The Presentation is dedicated to safety strategy of Bloc4 and Cluster D and focused on an analysis of risk conditions and on the real behaviour of safety valves in case of discharge. Protection of sub-atmospheric circuits is also discussed.

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)

Reactor containment purge and vent valve performance experiments

International Nuclear Information System (INIS)

Hunter, J.A.; Steele, R.; Watkins, J.C.

1985-01-01

Three nuclear-designed butterfly valves typical of those used in domestic nuclear power plant containment purge and vent applications were tested. For a comparison of responses, two eight-inch nominal pipe size valves with differing internal design were tested. For extrapolation insights, a 24-inch nominal pipe size valve was also tested. The valve experiments were performed with various piping configurations and valve disc orientations to the flow, to simulate various installation options in field application. As a standard for comparing the effects of the installation options, testing was also performed in a standard ANSI test section. Test cycles were performed at inlet pressures of 5 to 60 psig, while monitoring numerous test parameters, such as the valve disc position, valve shaft torque, mass flow rate, and the pressure and temperature at multiple locations throughout the test section. An experimental data base was developed to assist in the evaluation of the current analytical methods and to determine the influence of inlet pressure, inlet duct geometry, and valve orientation to the flow media on valve torque requirements, along with any resulting limitations to the extrapolation methods. 2 refs., 15 figs

Results of thermal test of metallic molybdenum disk target and fast-acting valve testing

Energy Technology Data Exchange (ETDEWEB)

Virgo, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Chemerisov, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Gromov, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Jonah, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Vandegrift, G. F. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

This report describes the irradiation conditions for thermal testing of helium-cooled metallic disk targets that was conducted on March 9, 2016, at the Argonne National Laboratory electron linac. The four disks in this irradiation were pressed and sintered by Oak Ridge National Laboratory from molybdenum metal powder. Two of those disks were instrumented with thermocouples. Also reported are results of testing a fast-acting-valve system, which was designed to protect the accelerator in case of a target-window failure.

Survey of industrial coal conversion equipment capabilities: valves

Energy Technology Data Exchange (ETDEWEB)

Bush, W. A.; Slade, E. C.

1978-06-01

A survey of the industrial capabilities of the valve and valve-actuator industry to supply large, high-pressure stop valves for the future coal conversion industry is presented in this report. Also discussed are development and testing capabilities of valve and valve-actuator manufacturers and anticipated lead times required to manufacture advanced design valves for the most stringent service applications. Results indicate that the valve and valve-actuator industry is capable of manufacturing in quantity equipment of the size and for the pressure and temperature ranges which would be required in the coal conversion industry. Valve manufacturers do not, however, have sufficient product application experience to predict the continuing functional ability of valves used for lock-hopper feeders, slurry feeders, and slag-char letdown service. Developmental and testing efforts to modify existing valve designs or to develop new valve concepts for these applications were estimated to range from 1 to 6 years. A testing facility to simulate actuation of critical valves under service conditions would be beneficial.

Enhancements to the Idaho National Engineering Laboratory motor-operated valve assessment software

International Nuclear Information System (INIS)

Holbrook, M.R.; Watkins, J.C.

1994-01-01

In January 1991, the U.S. Nuclear Regulatory Commission (USNRC) commenced Part 1 inspections to review licensee's motor-operated valve (MOV) programs that were developed to address Generic Letter 89-10, open-quotes Safety-Related Motor-Operated Valve Testing and Surveillanceclose quotes. In support, of this effort, the Isolation Valve Assessment (IVA) software, Version 3.10, was developed by the Idaho National Engineering Laboratory (INEL) to enable rapid in-depth review of MOV sizing and torque switch setting calculations. In 1994, the USNRC commenced Part 2 inspections, which involve a more in-depth review of MOV in situ testing relative to design-basis assumptions. The purpose of this paper is to describe the latest INEL and industry research that has been incorporated into Version 4.00 of the IVA software to support the latest round of inspections. Major improvements include (a) using dynamic and static test results to determine MOV performance parameters and validate design-basis engineering assumptions, (b) determining the stem/stem-nut coefficient of friction using new research-based techniques, (c) adding the ability to evaluate globe valves, and (d) incorporating new methods to account for the effects of high ambient temperature on the output torque of alternating current (ac) motors

Operating experience with sodium valves in the TNO-sodium test facilities

International Nuclear Information System (INIS)

Gasselt, M.L.G. van

1974-01-01

The development of sodium components for the SNR-300 in Holland has reached the stage where full scale testing in sodium has almost been finished and construction is at its height. It is against this background that a review is given of the weaknesses in one area or the other of the commercially available types of sodium valves used in TNO's smaller test facilities at Apeldoorn and TNO's 50 MW sodium components test facility at Hengelo. (U.S.)

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.)

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.

Thermal fatigue behavior of valves

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-08-15

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

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.

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

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.

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

International Nuclear Information System (INIS)

Haynes, H.D.

1990-01-01

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

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.)

Nuclear valves latest development

International Nuclear Information System (INIS)

Isaac, F.; Monier, M.

1993-01-01

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

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

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.

Operating experience and design criteria of sodium valves

International Nuclear Information System (INIS)

Markford, D.

1974-01-01

The information presented refers to sodium valve development for KNK and SNR-300 as well as for sodium test facilities on the INTERATOM site at Bensberg. Well in advance of KNK-I a number of sodium test facilities have been operated containing small and medium size valves of different design and manufacturer. The more stringent requirements for long range safe and reliable operation in KNK-I put forth a development program for the main primary and secondary circuit sodium valves. Operational experience gave rise to modification of the stem seal arrangement mainly, so KNK-II (which is the fast core for KNK reactor) will be run with modified sodium valves. Main pipe diameters in SNR-300 are in the range of 600 mm. Valve designs with rising shafts would require excessive space in the primary circuit cavities, therefore efforts have been directed towards introduction of different type valves. Due to the requirements of after-heat-removal a valve type with control capability had to be chosen. A special design of butterfly valves was selected for the primary and secondary circuits of SNR-300. The development and tests performed with this type of valve are described. In the field of small sodium valves, tests with a 50 mm diameter freeze-seal valve are reported, and the current status of bellows-seal-valves to be inserted into SNR-300 is discussed. (U.S.)

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

Valve stem packing seal test results for primary heat transport system conditions in Canadian nuclear generating stations

International Nuclear Information System (INIS)

Dixon, D.F.; Farrell, J.M.; Coutinho, R.F.

1978-06-01

Valve stem packing tests were done to obtain performance data on packing already in CANDU-PHW reactor service and on alternative packings. Most of the tests were replicated. Results are presented for ten packings tested under two stem cycle modes; leakage, packing consolidation and packing friction were the main responses. Packing tests were performed with water at close to CANDU-PHW reactor primary heat transport (PHT) system conditions (288 deg C and 10 MPa), but without ionizing radiation. The test rigs had rising, rotating stems. Stuffing box dimensions were typical of a standard Velan valve; packings were spring loaded to control applied packing stress

Valve packing manual. A maintenance application guide

International Nuclear Information System (INIS)

Aikin, J.A.; McCutcheon, R.G.; Cumming, D.

1997-01-01

Since 1970, AECL Chalk River Mechanical Equipment Development (MED) branch has invested over 175 person years in testing related to improving valve packing performance. Successful developments, including, 'live-loading', reduced packing heights, and performance-based packing qualification testing have been implemented. Since 1986, MED and the Integrated Valve Actuator Program Task Force - Valve Packing Steering Committee (IVAP-VPSC) have been involved in the development of combination die-formed graphite packing for use in CANDU plants. Many reports, articles, and specifications have been issued. Due to the large amount of test data and reports, a more user-friendly document has been prepared for everyday use. The Valve Packing Manual is based on many years of MED research and testing, as well as operating experience from CANDU nuclear generating stations (NGS). Since 1986, packing research and testing has been funded by the CANDU Owners Group (COG), the Electric Power Research Institute (EPRI), and participating valve packing manufacturers. The Valve Packing Manual (VPM) provides topical summaries of all work related to valve packing done since 1985. It includes advances in configuration design, stem packing friction, materials specifications, and installation procedures. This paper provides an overview on the application of the VPM with a focus on qualification testing, packing configuration, and stem packing friction. (author)

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

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

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 

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

Directory of Open Access Journals (Sweden)

Hwang Bae

2017-08-01

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

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

Experimental substantiation of the design of a prosthetic heart valve for «valve-in-valve» implantation

Directory of Open Access Journals (Sweden)

K. Yu. Klyshnikov

2017-01-01

Full Text Available The aim of the study was to perform a series of in vitro tests of a prototype of the developing heart valve prosthesis to evaluate its functional characteristics. Materials and methods. In this work we have used the frames and full prototypes of the prosthesis, consisting of a stent-like stainless steel support frame with mounted biological leaflets and cover. The authors evaluated the calculated and experimental forces necessary for the displacement of the sutureless implanted prosthesis using the test machine under uniaxial tension. The risk of defects and damages to the supporting framework as a result of implantation was evaluated by scanning electron microscopy. The hydrodynamic characteristics of the prosthesis were investigated under physiological conditions and «valvein-valve» implantation. Evaluation of the ergonomics and applicability of the proposed construction on the cadaver heart model of cattle was carried out. Results. As a result of the forces assessment, it was found that the force required to shear the prosthesis was 3.12 ± 0.37 N, while the calculated value was 1.7 N, which is significantly lower than the obtained value. The comparison of the images obtained with small and large magnifications demonstrated the absence of critical surface defects. Additional analysis under the super-large magnifications also did not reveal problem areas. During the hydrodynamic study, it was shown that the average transplant gradient increased slightly from 2.8–3.4 to 3.2–4.5 mm Hg for the initial prosthesis and the «valve-in-valve» complex, respectively. The decrease of the effective orifice area was 6–9% relative to the initial one. Evaluation of the implantation technique demonstrated the consistency of the approach: the use of the developed holder in combination with the balloon implantation system made it possible to position the prosthesis throughout the procedure. Conclusion. The series of tests demonstrates the consistency

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.)

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)

Performance testing of prototype live loaded packed stem seals for large gate valves in pressurized hot water

International Nuclear Information System (INIS)

Pothier, N.E.

1976-01-01

Prototype live loaded packed stem seals for large gate valves have been tested in a laboratory. The test fluid was demineralized water at 547 K, 8.27 MPa and pH 10. Nine packing configurations were tested; three different commercial brands of asbestos/graphite valve packings and three different sizes for each packing brand. Conventional and live loaded packed stem seals are briefly described. Stem leakage, packing consolidation and stem friction data are given. For all tests, leakage rates of less than 10 g d -1 were observed. It was also observed that stem friction was significantly affected by thermal expansion of the stem. (author)

Project W-314 specific test and evaluation plan 241-AN-B valve pit

International Nuclear Information System (INIS)

Hays, W.H.

1998-01-01

The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AN-B Valve Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation Plan (TEP)

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

Outlet strut fracture of Björk-Shiley convexo-concave valves: can valve-manufacturing characteristics explain the risk?

Science.gov (United States)

Omar, R Z; Morton, L S; Beirne, M; Blot, W J; Lawford, P V; Hose, R; Taylor, K M

2001-06-01

Björk-Shiley 60 degrees convexo-concave prosthetic heart valves (Shiley, Inc, Irvine, Calif, a subsidiary of Pfizer, Inc) continue to be a concern for approximately 35,000 nonexplanted patients worldwide, with approximately 600 events reported to the manufacturer to date. Fractures of the outlet struts of the valves began to appear in the early 1980s and have continued to the present, but their causes are only partially understood. A matched case-control study was conducted evaluating manufacturing records for 52 valves with outlet strut fractures and 248 control subjects matched for age at implantation, valve size, and valve position. In addition to the risk factors recognized as determinants of outlet strut fracture, the United Kingdom case-control study has observed 7- to 9-fold increased risk with performance of multiple hook deflection tests. This test was performed more than once, usually after rework on the valve. Six valves in this study underwent multiple hook deflection tests, of which 4 experienced an outlet strut fracture. Cracks and further rework were noted for these valves. Significant associations were also observed between outlet strut fracture and disc-to-strut gap measurements taken before the attachment of the sewing ring. It is our view that a combination of factors related to valve design, manufacturing process, and patient characteristics are responsible for outlet strut fractures of Björk-Shiley convexo-concave valves. Multiple hook deflection tests have emerged as a potential new risk factor for outlet strut fracture in both The Netherlands and the United Kingdom. This factor appears to be correlated with the presence of other abnormalities. A further study is needed to investigate the factors correlated with multiple hook deflection tests. On confirmation of risk, the presence of multiple hook deflection tests may be added to equations, quantifying the risk of outlet strut fracture for comparison against risk of mortality and serious

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.

Development of multilayer piezoelectric actuator valve for JT-60

International Nuclear Information System (INIS)

Miyo, Yasuhiko; Hiratsuka, Hajime; Masui, Hiroshi; Hosogane, Nobuyuki; Miya, Naoyuki

2001-11-01

In order to improve the gas injection valve used for the operation of JT-60, a new type of valve (multilayer piezoelectric actuator valve) was developed. The conventional valve (bimorph piezoelectric valve) has been used for 15 years since the beginning of experimental operation in April, 1985. However, it came to be hard to keep the performance of the valve because of the deterioration of the driving source, i.e. piezoelectric element. Developments of the new valve were carried out based on experiences through experimental operations in JT-60. Requirements for the design are: (1) to be hard structure for making a sheet leak, (2) to allow a repair work at atmosphere side without an air vent of the vacuum vessel, (3) to be more smaller and lighter compared with the conventional one, and (4) to have a high maintenance efficiency by utilizing of the commercial piezoelectric elements and power supplies. The newly developed valve was examined with various tests such as gas flow characteristic test, high magnetic field proof test, high temperature proof test and gas flow rate test for aged deterioration. Results, confirm that the performance of the valve is applicable for JT-60 operations. (author)

Current status of nuclear safety research

International Nuclear Information System (INIS)

Anon.

1977-01-01

Efforts at nuclear safety research have expanded year by year in Japan, in term of money and technical achievement. The Atomic Energy Commission set last year the five year nuclear safety research program, a guideline by which various research institutes will be able to develop their own efforts in a concerted manner. From the results of the nuclear safety research which cover very wide areas ranging from reactor engineering safety, safety of nuclear fuel cycle facilities, prevention of radiation hazards to the adequate treatment and disposal of radioactive wastes, AIJ hereafter focuses of LWR engineering safety and prevents two articles, one introducing the current results of the NSSR program developed by JAERI and the other reporting the LWR reliability demonstration testing projects being promoted by MITI. The outline of these demonstration tests was reported in this report. The tests consist of earthquake resistance reliability test of nuclear power plants, steam generator reliability tests, valve integrity tests, fuel assembly reliability tests, reliability tests of heat affected zones and reliability tests of pumps. (Kobatake, H.)

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

Investigations on pneumatically forced-actuated compressor valves

Science.gov (United States)

Stöckel, Christian; Thomas, Christiane; Nickl, Jörg; Hesse, Ullrich

2017-08-01

In the present paper the performance of a novel designed valve for reciprocating piston machines is investigated, which makes existing compressors utilizable for operating as expander. Three design parameters were identified as critical for the valves performance particularly in forced actuated mode. Within a numerical simulation a study on the crucial geometrical parameters, the influence could be observed. Afterwards the experimental setup for the integral test of the valve design is presented and also additional tests for single valve components.

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

Characteristic analysis and experiment of pneumatic servo valve

International Nuclear Information System (INIS)

Kim, Dong Soo; Lee, Won Hee; Choi, Byung Oh

2004-01-01

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

Evaluation of SPACE code for simulation of inadvertent opening of spray valve in Shin Kori unit 1

International Nuclear Information System (INIS)

Kim, Seyun; Youn, Bumsoo

2013-01-01

SPACE code is expected to be applied to the safety analysis for LOCA (Loss of Coolant Accident) and Non-LOCA scenarios. SPACE code solves two-fluid, three-field governing equations and programmed with C++ computer language using object-oriented concepts. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code. To evaluate the analysis capability for the transient phenomena in the actual nuclear power plant, an inadvertent opening of spray valve in startup test phase of Shin Kori unit 1 was simulated with SPACE code

Hardfacing and packings for improved valve performance

International Nuclear Information System (INIS)

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

2003-01-01

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

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.

Analysis of valve failures from the NUCLARR data base

International Nuclear Information System (INIS)

Moore, L.M.

1997-11-01

The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) contains data on component failures with categorical and qualifying information such as component design, normal operating state, system application and safety grade information which is important to the development of risk-based component surveillance testing requirements. This report presents descriptions and results of analyses of valve component failure data and covariate information available in the document Nuclear Computerized Library for Assessing Reactor Reliability Data Manual, Part 3: Hardware Component Failure Data (NUCLARR Data Manual). Although there are substantial records on valve performance, there are many categories of the corresponding descriptors and qualifying information for which specific values are missing. Consequently, this limits the data available for analysis of covariate effects. This report presents cross tabulations by different covariate categories and limited modeling of covariate effects for data subsets with substantive non-missing covariate information

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

Check valves aging assessment

International Nuclear Information System (INIS)

Haynes, H.D.

1991-01-01

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

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.

Low noise control valve

International Nuclear Information System (INIS)

Christie, R.S.

1975-01-01

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

Study of typical nuclear containment purge valves in an accident environment

International Nuclear Information System (INIS)

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

1986-08-01

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

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.

Estimation of the common cause failure probabilities of the components under mixed testing schemes

International Nuclear Information System (INIS)

Kang, Dae Il; Hwang, Mee Jeong; Han, Sang Hoon

2009-01-01

For the case where trains or channels of standby safety systems consisting of more than two redundant components are tested in a staggered manner, the standby safety components within a train can be tested simultaneously or consecutively. In this case, mixed testing schemes, staggered and non-staggered testing schemes, are used for testing the components. Approximate formulas, based on the basic parameter method, were developed for the estimation of the common cause failure (CCF) probabilities of the components under mixed testing schemes. The developed formulas were applied to the four redundant check valves of the auxiliary feed water system as a demonstration study for their appropriateness. For a comparison, we estimated the CCF probabilities of the four redundant check valves for the mixed, staggered, and non-staggered testing schemes. The CCF probabilities of the four redundant check valves for the mixed testing schemes were estimated to be higher than those for the staggered testing scheme, and lower than those for the non-staggered testing scheme.

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.

Project W-314 specific test and evaluation plan for 241-AN-A valve pit

International Nuclear Information System (INIS)

Hays, W.H.

1998-01-01

The purpose of this Specific Test and Evaluation Plan (STEP) is to provide a detailed written plan for the systematic testing of modifications made to the 241-AN-A Valve Pit by the W-314 Project. The STEP develops the outline for test procedures that verify the system's performance to the established Project design criteria. The STEP is a lower tier document based on the W-314 Test and Evaluation Plan (TEP)

The Development of Computer Code for Safety Injection Tank (SIT) with Fluidic Device(FD) Blowdown Test

International Nuclear Information System (INIS)

Lee, Joo Hee; Kim, Tae Han; Choi, Hae Yun; Lee, Kwang Won; Chung, Chang Kyu

2007-01-01

Safety Injection Tanks (SITs) with the Fluidic Device (FD) of APR1400 provides a means of rapid reflooding of the core following a large break Loss Of Coolant Accident (LOCA), and keeping it covered until flow from the Safety Injection Pump (SIP) becomes available. A passive FD can provide two operation stages of a safety water injection into the RCS and allow more effective use of borated water in case of LOCA. Once a large break LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced to a lower flow rate. The conventional computer code 'TURTLE' used to simulate the blowdown of OPR1000 SIT can not be directly applied to simulate a blowdown process of the SIT with FD. A new computer code is needed to be developed for the blowdown test evaluation of the APR1400 SIT with FD. Korea Power Engineering Company (KOPEC) has developed a new computer code to analyze the characteristics of the SIT with FD and validated the code through the comparison of the calculation results with the test results obtained by Ulchin 5 and 6 units pre-operational test and VAlve Performance Evaluation Rig (VAPER) tests performed by The Korea Atomic Energy Research Institute (KAERI)

Preliminary assessment of valve IST effectiveness

International Nuclear Information System (INIS)

Grove, E.; DiBiasio, A.; Carbonaro, J.

1994-01-01

A preliminary review of inservice testing effectiveness for Code Class 1,2, and 3 valves at nuclear power plants was performed. These requirements are specified by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, and the Operations and Maintenance Standard. The Institute of Nuclear Power Operations Nuclear Plant Reliability Data System (NPRDS) database was used to provide failure reports for these components for 1988 to 1992. This time period coincides with the issuance of Generic Letter 89-04, which resulted in a more consistent application of the requirements by the licensees. For this time period, 8,593 valve failures were identified. From the review of the NPRDS database, the primary failure causes and failure modes for motor-operated valves (MOV), air-operated valves (AOV), and check valves (CV) were identified. Solenoid-operated valves were not reviewed in this study. Plant testing programs were effective in identifying approximately 60% of the CV failures, 46% of the AOV failures, and 44% of the MOV failures

Cardiopulmonary Exercise Testing in Patients with Asymptomatic or Equivocal Symptomatic Aortic Stenosis: Feasibility, Reproducibility, Safety and Information Obtained on Exercise Physiology.

Science.gov (United States)

van Le, Douet; Jensen, Gunnar Vagn Hagemann; Carstensen, Steen; Kjøller-Hansen, Lars

2016-01-01

The aim of this study was to determine the feasibility, reproducibility, safety and information obtained on exercise physiology from cardiopulmonary exercise testing (CPX) in patients with aortic stenosis. Patients with an aortic valve area (AVA) exercise, lower peak heart rate and FEV1, and higher VE/VCO2, but not by AVA index. Equivocal symptomatic status and a low gradient but high valvulo-arterial impedance were associated with a lower pVO2, but not with an inability to increase stroke volume. In total, 18 patients were referred for valve replacement. At 1 year, no cardiovascular deaths had occurred. CPX was feasible and reproducible and provided comprehensive data on exercise physiology. A CPX-guided treatment strategy was safe up to 1 year. © 2015 S. Karger AG, Basel.

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...

A microfluidic device with multi-valves system to enable several simultaneous exposure tests on Caenorhabditis elegans

International Nuclear Information System (INIS)

Jung, Jaehoon; Masaru, Takeuchi; Nakajima, Masahiro; Huang, Qiang; Fukuda, Toshio

2014-01-01

In this paper, we report on a microfluidic device with a multi-valve system to conduct several exposure tests on Caenorhabditis elegans (C. elegans) simultaneously. It has pneumatic valves and no-moving-parts (NMP) valves. An NMP valve is incorporated with a chamber and enables the unidirectional movement of C. elegans in the chamber; once worms are loaded into the chamber, they cannot exit, regardless of the flow direction. To demonstrate the ability of the NMP valve to handle worms, we made a microfluidic device with three chambers. Each chamber was used to expose worms to Cd and Cu solutions, and K-medium. A pair of electrodes was installed in the device and the capacitance in-between the electrode was measured. When a C. elegans passed through the electrodes, the capacitance was changed. The capacitance change was proportional to the body volume of the worm, thus the body volume change by the heavy metal exposure was measured in the device. Thirty worms were divided into three groups and exposed to each solution. We confirmed that the different solutions induced differences in the capacitance changes for each group. These results indicate that our device is a viable method for simultaneously analyzing the effect of multiple stimuli on C. elegans. (paper)

Automatic fire hydrant valve development

International Nuclear Information System (INIS)

Drumheller, K.

1976-01-01

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

ASME XI stroke time testing of solenoid valves at Connecticut Yankee Station

Energy Technology Data Exchange (ETDEWEB)

Martin, C.W.

1996-12-01

Connecticut Yankee Atomic Power Company has developed the capability of measuring the stroke times of AC and DC solenoid valves. This allows the station to measure the stroke time of any solenoid valve in the plant, even those valves which do not have valve stem position indicators. Connecticut Yankee has adapted the ITI MOVATS Checkmate 3 system, using a signal input from a Bruel and Kjaer (B&K) Model 4382 acoustic accelerometer and the Schaumberg Campbell Associates (SCA) Model SCA-1148 dual sensor, which is a combined accelerometer and gaussmeter.

ASME XI stroke time testing of solenoid valves at Connecticut Yankee Station

International Nuclear Information System (INIS)

Martin, C.W.

1996-01-01

Connecticut Yankee Atomic Power Company has developed the capability of measuring the stroke times of AC and DC solenoid valves. This allows the station to measure the stroke time of any solenoid valve in the plant, even those valves which do not have valve stem position indicators. Connecticut Yankee has adapted the ITI MOVATS Checkmate 3 system, using a signal input from a Bruel and Kjaer (B ampersand K) Model 4382 acoustic accelerometer and the Schaumberg Campbell Associates (SCA) Model SCA-1148 dual sensor, which is a combined accelerometer and gaussmeter

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)

Evaluating All-Metal Valves for Use in a Tritium Environment

Energy Technology Data Exchange (ETDEWEB)

Houk, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Payton, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-09-28

In the tritium gas processing system, it is desired to minimize polymer components due to their degradation from tritium exposure (beta decay). One source of polymers in the tritium process is valve components. A vendor has been identified that manufactures a valve that is marketed as being made from all-metal construction. This manufacturer, Ham-Let Group, manufactures a diaphragm valve (3LE series) that claims to be made entirely of metal. SRNL procured twelve (12) Ham-Let diaphragm valves for characterization and evaluation. The characterization tests include identification of the maximum pressure of these valves by performing pressure and burst tests. Leak tests were performed to ensure the valves do not exceed the acceptable leak rate for tritium service. These valves were then cycled in a nitrogen gas and/or vacuum environment to ensure they would be durable in a process environment. They were subsequently leak tested per ASTM protocol to ensure that the valves maintained their leak tight integrity. A detailed material analysis was also conducted to determine hydrogen and tritium compatibility.

Flow oscillations on the steam control valve in the middle opening condition. Clarification of the effects of valve body and valve seat by steam experiments

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio

2007-01-01

A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop the countermeasures by CFD (Computational Fluid Dynamics) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the new valve shape (named 'Extended Valve') that can suppress the pressure fluctuations by air experiments and CFD calculations. Then, we also conducted steam experiments and CFD calculations to understand the differences between air and the steam, and found that the pressure fluctuations in the middle opening condition also occurred in the steam tests and the differences between the air and steam were not remarkable. In this report, to clarify the effects of valve and valve seat shape in steam flow condition, we conduct the steam experiments with various valve and seat shape. As a result, we find the change of the valve seat can decrease the amplitude of pressure fluctuations, but can not quite suppress the pressure fluctuations in the middle opening condition. Then, we apply the 'Extended Valve' to clarify the valve shape effect, and find that the extended valve suppresses the pressure fluctuations in the middle opening condition completely and decreases the pressure amplitude drastically. (author)

Björk-Shiley convexoconcave valves: susceptibility artifacts at brain MR imaging and mechanical valve fractures.

Science.gov (United States)

van Gorp, Maarten J; van der Graaf, Yolanda; de Mol, Bas A J M; Bakker, Chris J G; Witkamp, Theo D; Ramos, Lino M P; Mali, Willem P T M

2004-03-01

To assess the relationship between heart valve history and susceptibility artifacts at magnetic resonance (MR) imaging of the brain in patients with Björk-Shiley convexoconcave (BSCC) valves. MR images of the brain were obtained in 58 patients with prosthetic heart valves: 20 patients had BSCC valve replacements, and 38 had other types of heart valves. Two experienced neuroradiologists determined the presence or absence of susceptibility artifacts in a consensus reading. Artifacts were defined as characteristic black spots that were visible on T2*-weighted gradient-echo MR images. The statuses of the 20 explanted BSCC valves-specifically, whether they were intact or had an outlet strut fracture (OSF) or a single-leg fracture (SLF)-had been determined earlier. Number of artifacts seen at brain MR imaging was correlated with explanted valve status, and differences were analyzed with nonparametric statistical tests. Significantly more patients with BSCC valves (17 [85%] of 20 patients) than patients with other types of prosthetic valves (18 [47%] of 38 patients) had susceptibility artifacts at MR imaging (P =.005). BSCC valve OSFs were associated with a significantly higher number of artifacts than were intact BSCC valves (P =.01). No significant relationship between SLF and number of artifacts was observed. Susceptibility artifacts at brain MR imaging are not restricted to patients with BSCC valves. These artifacts can be seen on images obtained in patients with various other types of fractured and intact prosthetic heart valves. Copyright RSNA, 2004

Flow induced vibration of the large-sized sodium valve for MONJU

Energy Technology Data Exchange (ETDEWEB)

Sato, K [Sodium Engineering Division, O-arai Engineering Centre, Power Reactor and Nuclear Fuel Development Corporation, Nariata-cho, O-arai Machi, Ibaraki-ken (Japan)

1977-12-01

Measurements have been made on the hydraulic characteristics of the large-sized sodium valves in the hydraulic simulation test loop with water as fluid. The following three prototype sodium valves were tested; (1) 22-inch wedge gate type isolation valve, (2) 22-inch butterfly type isolation valve, and (3) 16-inch butterfly type control valve. In the test, accelerations of flow induced vibrations were measured as a function of flow velocity and disk position. The excitation mechanism of the vibrations is not fully interpreted in these tests due to the complexity of the phenomena, but the experimental results suggest that it closely depends on random pressure fluctuations near the valve disk and flow separation at the contracted cross section between the valve seat and the disk. The intensity of flow induced vibrations suddenly increases at a certain critical condition, which depends on the type of valve and is proportional to fluid velocity. (author)

Flow induced vibration of the large-sized sodium valve for MONJU

International Nuclear Information System (INIS)

Sato, K.

1977-01-01

Measurements have been made on the hydraulic characteristics of the large-sized sodium valves in the hydraulic simulation test loop with water as fluid. The following three prototype sodium valves were tested; (1) 22-inch wedge gate type isolation valve, (2) 22-inch butterfly type isolation valve, and (3) 16-inch butterfly type control valve. In the test, accelerations of flow induced vibrations were measured as a function of flow velocity and disk position. The excitation mechanism of the vibrations is not fully interpreted in these tests due to the complexity of the phenomena, but the experimental results suggest that it closely depends on random pressure fluctuations near the valve disk and flow separation at the contracted cross section between the valve seat and the disk. The intensity of flow induced vibrations suddenly increases at a certain critical condition, which depends on the type of valve and is proportional to fluid velocity. (author)

Development of magnetic drive packless valves for commercial purpose

Energy Technology Data Exchange (ETDEWEB)

Hwang, Seong Tae; Choi, Yoon Dong; Jeong, Ji Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

A study on development of magnetic drive packless valves for commercial purpose showed the results as follows; A. Maximization of torque (1) Arranging method of magnets (2) Effects on the distance between magnetic holders (3) Improvement of the valve disc for the torque maximization. B. Study on the high pressure structure of the valve (1) Analysis of the pressure-resisting structure of the magnetic power transmitting device (2) Calculation model and program (a) Calculation model (b) Program used in calculation (3) Structure analysis and verification test for the pressurized plate. C. Study on the valve compacting. D. Technical specification for the valve manufacture. E. Inspection and test method for valve. 10 tabs., 10 figs. (Author).

Development of magnetic drive packless valves for commercial purpose

International Nuclear Information System (INIS)

Hwang, Seong Tae; Choi, Yoon Dong; Jeong, Ji Young

1995-12-01

A study on development of magnetic drive packless valves for commercial purpose showed the results as follows; A. Maximization of torque (1) Arranging method of magnets (2) Effects on the distance between magnetic holders (3) Improvement of the valve disc for the torque maximization. B. Study on the high pressure structure of the valve (1) Analysis of the pressure-resisting structure of the magnetic power transmitting device (2) Calculation model and program (a) Calculation model (b) Program used in calculation (3) Structure analysis and verification test for the pressurized plate. C. Study on the valve compacting. D. Technical specification for the valve manufacture. E. Inspection and test method for valve. 10 tabs., 10 figs. (Author)

Leak-thight seals got high pressure testing of pipes, tanks, valves

International Nuclear Information System (INIS)

Estrade, J.

1985-01-01

Leak-tight seals ensure quick, safe and efficient testing of pipes with plain-ended or flanged openings, valves with flanged or welded edges, manifields, recipients, etc. They are inserted into the pipe end manually then simply a slight turn of the seal treated wheel commences the pressure test. Hydraulic pressure is supplied by a pump through the inlet seal and air is purged through the outlet seal which then closes. The higher the pressure, the greater the sealing strength of the seal which prevents accidental unplugging. There are different types of seals: for interior plain-ended openings, for pipes with plain-ended opening, for flanged pipes. (author)

Practical use of valve seating machine with remote control system for main steam isolation valve at N.P.S

International Nuclear Information System (INIS)

Ito, Sadao; Noda, Hiroshi; Sadamura, Morito; Utsunomiya, Yasushi.

1975-01-01

The main steam isolation valves in BWR power stations are installed at the boundary of reactor containment vessels, and 2 valves in each main steam system total 8 valves in a plant. They are pneumatically operated Y type globe valves for preventing the release of radioactive substances in the atmosphere in case of the breaking of main steam pipes and also preventing the loss of coolant in case of the breaking of recirculating equipments. Therefore careful leak test, inspection, and seat-fitting are carried out to the valves at each regular maintenance. The manual maintenance work is difficult because of narrow space and the reduction of exposure, and the seat-fitting work requires the skill of high degree, therefore Okano Valve Manufacturing Co. and Tokyo Electric Power Co. jointly started the research and development of an automatic valve seating machine, and successfully put it to practical use in Fukushima No.1 Nuclear Power Station in Nov. 1974. First, the problems in the manual seat-fitting work were investigated, and the means to mechanically solve them were materialized with a prototype machine. After its mock-up test, an actual machine was designed and manufactured. The test result showed remarkable reduction of exposure and labor-saving, and the leak evaluation was sufficiently below the allowable value. (Kako, I.)

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

Application of new designed butterfly type intermediate valve for nuclear steam turbine

International Nuclear Information System (INIS)

Matsumura, Kazuhiro; Kawamata, Susumu; Fujita, Isao; Taketomo, Seiki.

1991-01-01

To cope with a large capacity nuclear steam turbine, a butterfly type intermediate valve has been developed. Compared to the conventional valve, or globe valve, the butterfly valve has the following design features: a) Higher thermal efficiency due to lower pressure loss, b) Easier maintenance due to simplified construction, and c) Lower station cost due to the smaller size of the valve assembly. An experiment with a scaled-down test valve was carried out using compressed air. Subsequently a full-scale valve was tested using steam under actual steam conditions. As a result, these tests gave us no problems. The first nuclear turbine (1100MW) equipped with a butterfly valve is operating satisfactorily with good performance as expected. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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.

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)

Development of an effective valve packing program

Energy Technology Data Exchange (ETDEWEB)

Hart, K.A.

1996-12-01

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

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

9 CFR 113.39 - Cat safety tests.

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Cat safety tests. 113.39 Section 113... Procedures § 113.39 Cat safety tests. The safety tests provided in this section shall be conducted when... recommended for use in cats. (a) The cat safety test provided in this paragraph shall be used when the Master...

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

Improvement of a Pneumatic Control Valve with Self-Holding Function

Science.gov (United States)

Dohta, Shujiro; Akagi, Tetsuya; Kobayashi, Wataru; Shimooka, So; Masago, Yusuke

2017-10-01

The purpose of this study is to develop a small-sized, lightweight and low-cost control valve with low energy consumption and to apply it to the assistive system. We have developed some control valves; a tiny on/off valve using a vibration motor, and an on/off valve with self-holding function. We have also proposed and tested the digital servo valve with self-holding function using permanent magnets and a small-sized servo motor. In this paper, in order to improve the valve, an analytical model of the digital servo valve is proposed. And the simulated results by using the analytical model and identified parameters were compared with the experimental results. Then, the improved digital servo valve was designed based on the calculated results and tested. As a result, we realized the digital servo valve that can control the flow rate more precisely while maintaining its volume and weight compared with the previous valve. As an application of the improved valve, a position control system of rubber artificial muscle was built and the position control was performed successfully.

Aperture Valve for the Mars Organic Molecule Analyzer (MOMA)

Science.gov (United States)

Hakun, Claef F.; Engler, Charles D.; Barber, Willie E.; Canham, John S.

2014-01-01

NASA's participation in the multi-nation ExoMars 2018 Rover mission includes a critical astrobiology Mass Spectrometer Instrument on the Rover called the Mars Organic Molecule Analyzer (MOMA). The Aperture Valve is a critical electromechanical valve used by the Mass Spectrometer to facilitate the transfer of ions from Martian soil to the Mass Spectrometer for analysis. The MOMA Aperture Valve development program will be discussed in terms of the Initial valve design and subsequent improvements that resulted from prototype testing. The Initial Aperture Valve concept seemed promising, based on calculations and perceived merits. However, performance results of this design were disappointing, due to delamination of TiN and DLC coatings applied to the Titanium base metals, causing debris from the coatings to seize the valve. While peer reviews and design trade studies are important forums to vet a concept design, results from testing should not be underestimated.Despite the lack of development progress to meet requirements, valuable information from weakness discovered in the Initial Valve design was used to develop a second, more robust Aperture valve. Based on a check-ball design, the ETU flight valve design resulted in significantly less surface area to create the seal. Moreover, PVD coatings were eliminated in favor of hardened, nonmagnetic corrosion resistant alloys. Test results were impressive, with the valve achieving five orders of magnitude better sealing leak rate over end of life requirements. Cycle life was equally impressive, achieving 280,000 cycles without failure.

BWR reactor water cleanup system flexible wedge gate isolation valve qualification and high energy flow interruption test

International Nuclear Information System (INIS)

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

1989-10-01

This report presents the results of research performed to develop technical insights for the NRC effort regarding Generic Issue 87, ''Failure of HPCI Steam Line Without Isolation.'' Volume III of this report contains the data and findings from the original research performed to assess the qualification of the valves and reported in EGG-SSRE-7387, ''Qualification of Valve Assemblies in High Energy BWR Systems Penetrating Containment.'' We present the original work here to complete the documentation trail. The recommendations contained in Volume III of this report resulted in the test program described in Volume I and II. The research began with a survey to characterize the population of normally open containment isolation valves in those process lines that connect to the primary system and penetrate containment. The qualification methodology used by the various manufacturers identified in the survey is reviewed and deficiencies in that methodology are identified. Recommendations for expanding the qualification of valve assemblies for high energy pipe break conditions are presented. 11 refs., 1 fig., 2 tabs

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.