WorldWideScience

Sample records for safety test pressure

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-10-15

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

  2. Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

    International Nuclear Information System (INIS)

    Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yoon, Eun-Koo; Shin, Yong-Cheol; Min, Kyoung-Ho; Park, Jong-Kuk; Choi, Nam-Hyun; Bang, Yun-Gon; Seo, Chan-Jong; Yi, Sung-Jae; Park, Hyun-Sik

    2016-01-01

    SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors

  3. Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

    SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors.

  4. Safety demonstration tests on pressure rise in ventilation system and blower integrity of a fuel-reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Takada, Junichi; Suzuki, Motoe; Tsukamoto, Michio; Koike, Tadao; Nishio, Gunji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1996-12-01

    In JAERI, the demonstration test was carried out as a part of safety researches of the fuel-reprocessing plant using a large-scale facility consist of cells, ducts, dumpers, HEPA filters and a blower, when an explosive burning due to a rapid reaction of thermal decomposition for solvent/nitric acid occurs in a cell of the reprocessing plant. In the demonstration test, pressure response propagating through the facility was measured under a blowing of air from a pressurized tank into the cell in the facility to elucidate an influence of pressure rise in the ventilation system. Consequently, effective pressure decrease in the facility was given by a configuration of cells and ducts in the facility. In the test, transient responses of HEPA filters and the blower by the blowing of air were also measured to confirm the integrity. So that, it is confirmed that HEPA filters and the blower under pressure loading were sufficient to maintain the integrity. The content described in this report will contribute to safety assessment of the ventilation system in the event of explosive burning in the reprocessing plant. (author)

  5. Guide On Safety Tests

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-05-15

    This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

  6. Guide On Safety Tests

    International Nuclear Information System (INIS)

    1987-05-01

    This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

  7. Pressure Safety Orientation Live #769

    Energy Technology Data Exchange (ETDEWEB)

    Glass, George [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-17

    Pressure Safety Orientation (course #769) introduces workers at Los Alamos National Laboratory (LANL) to the Laboratory Pressure Safety Program and to pressure-related hazards. This course also affords a hands-on exercise involving the assembly of a simple pressure system. This course is required for all LANL personnel who work on or near pressure systems and are exposed to pressure-related hazards. These personnel include pressure-system engineers, designers, fabricators, installers, operators, inspectors, maintainers, and others who work with pressurized fluids and may be exposed to pressure-related hazards.

  8. Application of ultrasonic testing technique to detect gas accumulation in important pipings for pressurized water reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Fushimi, Yasuyuki [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    Since 1988, the USNRC has pointed out that gas-binding events might occur at high head safety injection (HHSI) pumps of pressurized water reactors (PWRs). In Japanese PWR plants, corrective actions were taken in response to gas-binding events that occurred on HHSI pumps in the USA, so no gas accumulation event has been reported so far. However, when venting frequency is prolonged with operating cycle extension, the probability of gas accumulation in pipings may increase as in the USA. The purpose of this study was to establish a technique to identify gas accumulation and to measure the gas volume accurately. Taking dominant causes of the gas-binding events in the USA into consideration, we pointed out the following sections in the Japanese PWRs where gas srtipping and/or gas accumulation might occur: residual heat removal system pipings and charging/safety injection pump minimum flow line. Then an ultrasonic testing technique, adopted to identify gas accumulation in the USA, was applied to those sections of the typical Japanese PWR. Consequently, no gas accumulation was found in those pipings. (author)

  9. Pressure Safety Program Implementation at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Lower, Mark [ORNL; Etheridge, Tom [ORNL; Oland, C. Barry [XCEL Engineering, Inc.

    2013-01-01

    The Oak Ridge National Laboratory (ORNL) is a US Department of Energy (DOE) facility that is managed by UT-Battelle, LLC. In February 2006, DOE promulgated worker safety and health regulations to govern contractor activities at DOE sites. These regulations, which are provided in 10 CFR 851, Worker Safety and Health Program, establish requirements for worker safety and health program that reduce or prevent occupational injuries, illnesses, and accidental losses by providing DOE contractors and their workers with safe and healthful workplaces at DOE sites. The regulations state that contractors must achieve compliance no later than May 25, 2007. According to 10 CFR 851, Subpart C, Specific Program Requirements, contractors must have a structured approach to their worker safety and health programs that at a minimum includes provisions for pressure safety. In implementing the structured approach for pressure safety, contractors must establish safety policies and procedures to ensure that pressure systems are designed, fabricated, tested, inspected, maintained, repaired, and operated by trained, qualified personnel in accordance with applicable sound engineering principles. In addition, contractors must ensure that all pressure vessels, boilers, air receivers, and supporting piping systems conform to (1) applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (2004) Sections I through XII, including applicable code cases; (2) applicable ASME B31 piping codes; and (3) the strictest applicable state and local codes. When national consensus codes are not applicable because of pressure range, vessel geometry, use of special materials, etc., contractors must implement measures to provide equivalent protection and ensure a level of safety greater than or equal to the level of protection afforded by the ASME or applicable state or local codes. This report documents the work performed to address legacy pressure vessel deficiencies and comply

  10. Seismic assessment of safety-related structures: laboratory testing of the pressure relief duct frame at pickering NPP

    International Nuclear Information System (INIS)

    Ghobarah, A.; Biddah, A.; Pilette, C.

    1995-01-01

    The pressure relief duct (PRD) is a Special safety System in the CANDU-PHW multi-unit nuclear power plants (NPP). It is designed to contain and direct the outflow from the reactor building to the pressure suppression and containing systems in the vacuum building. The PRD is a large elevated reinforced concrete box structure of internal width of 6.1 m, height of 7.6 m, and wall thickness of 0.6 m. The PRD is 662 m long and is supported every 22 m by concrete frames of height of 21 m. Typical frame members are 1.8 m in depth and width. A representative elevation of the frame is presented. The section of the PRD under investigation was designed and constructed before the current seismic design codes were in effect. An assessment of the PRD structure when subjected to various levels of ground motion has shown that the frame has a limited seismic withstand capacity. Its seismic performance is dependent on the ductility of the beams and on the ability of the beam-column joint to transfer bending moments and shear forces. The objectives of this study are to provide the data to validate the frame analysis results through laboratory testing of a scaled specimen of the beam-column joint, and to compare the observed response with the response of a beam-column joint when the shear reinforcement is detailed according to current seismic design codes. (author). 3 refs., 10 figs

  11. FOOD SAFETY TESTING LABORATORY

    Data.gov (United States)

    Federal Laboratory Consortium — This laboratory develops screening assays, tests and modifies biosensor equipment, and optimizes food safety testing protocols for the military and civilian sector...

  12. Pressure locking test results

    Energy Technology Data Exchange (ETDEWEB)

    DeWall, K.G.; Watkins, J.C.; McKellar, M.G.; Bramwell, D. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1996-12-01

    The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, is funding the Idaho National Engineering Laboratory (INEL) in performing research to provide technical input for their use in evaluating responses to Generic Letter 95-07, {open_quotes}Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves.{close_quotes} Pressure locking and thermal binding are phenomena that make a closed gate valve difficult to open. This paper discusses only the pressure locking phenomenon in a flexible-wedge gate valve; the authors will publish the results of their thermal binding research at a later date. Pressure locking can occur when operating sequences or temperature changes cause the pressure of the fluid in the bonnet (and, in most valves, between the discs) to be higher than the pressure on the upstream and downstream sides of the disc assembly. This high fluid pressure presses the discs against both seats, making the disc assembly harder to unseat than anticipated by the typical design calculations, which generally consider friction at only one of the two disc/seat interfaces. The high pressure of the bonnet fluid also changes the pressure distribution around the disc in a way that can further contribute to the unseating load. If the combined loads associated with pressure locking are very high, the actuator might not have the capacity to open the valve. The results of the NRC/INEL research discussed in this paper show that the relationship between bonnet pressure and pressure locking stem loads appears linear. The results also show that for this valve, seat leakage affects the bonnet pressurization rate when the valve is subjected to thermally induced pressure locking conditions.

  13. Pressure locking test results

    International Nuclear Information System (INIS)

    DeWall, K.G.; Watkins, J.C.; McKellar, M.G.; Bramwell, D.

    1996-01-01

    The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, is funding the Idaho National Engineering Laboratory (INEL) in performing research to provide technical input for their use in evaluating responses to Generic Letter 95-07, open-quotes Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves.close quotes Pressure locking and thermal binding are phenomena that make a closed gate valve difficult to open. This paper discusses only the pressure locking phenomenon in a flexible-wedge gate valve; we will publish the results of our thermal binding research at a later date. Pressure locking can occur when operating sequences or temperature changes cause the pressure of the fluid in the bonnet (and, in most valves, between the discs) to be higher than the pressure on the upstream and downstream sides of the disc assembly. This high fluid pressure presses the discs against both seats, making the disc assembly harder to unseat than anticipated by the typical design calculations, which generally consider friction at only one of the two disc/seat interfaces. The high pressure of the bonnet fluid also changes the pressure distribution around the disc in a way that can further contribute to the unseating load. If the combined loads associated with pressure locking are very high, the actuator might not have the capacity to open the valve. The results of the NRC/INEL research discussed in this paper show that the relationship between bonnet pressure and pressure locking stem loads appears linear. The results also show that for this valve, seat leakage affects the bonnet pressurization rate when the valve is subjected to thermally induced pressure locking conditions

  14. 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.

  15. Safety of steel vessel Magnox pressure circuits

    International Nuclear Information System (INIS)

    Stokoe, T.Y.; Bolton, C.J.; Heffer, P.J.H.

    1991-01-01

    The maintenance of pressure circuit integrity is fundamental to nuclear safety at the steel vessel Magnox stations. To confirm continued pressure circuit integrity the CEGB, as part of the Long Term Safety Review, has carried out extensive assessment and inspection in recent years. The assessment methods and inspection techniques employed are based on the most modern available. Reactor pressure vessel integrity is confirmed by a combination of arguments including safety factors inferred from the successful pre-service overpressure test, leak-before-break analysis and probabilistic assessment. In the case of other parts of the pressure circuits that are more accessible, comprising the boiler shells and interconnecting gas duct work, in-service inspection is a major element of the safety substantiation. The assessment and inspection techniques and the materials property data have been underpinned for many years by extensive research and development programmes and in-reactor monitoring of representative samples has also been undertaken. The paper summarises the work carried out to demonstrate the long term integrity of the Magnox pressure circuits and provides examples of the results obtained. (author)

  16. LOFT pressurizer safety: relief valve reliability

    International Nuclear Information System (INIS)

    Brown, E.S.

    1978-01-01

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

  17. LOFT pressurizer safety: relief valve reliability

    Energy Technology Data Exchange (ETDEWEB)

    Brown, E.S.

    1978-01-18

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

  18. Blood Pressure Test

    Science.gov (United States)

    ... pressure monitors may have some limitations. Tracking your blood pressure readings It can be helpful in diagnosing or ... more Stage 2 high blood pressure (hypertension) Elevated blood pressure and stages 1 and 2 high blood pressure ( ...

  19. A Risk Analysis Methodology to Address Human and Organizational Factors in Offshore Drilling Safety: With an Emphasis on Negative Pressure Test

    Science.gov (United States)

    Tabibzadeh, Maryam

    According to the final Presidential National Commission report on the BP Deepwater Horizon (DWH) blowout, there is need to "integrate more sophisticated risk assessment and risk management practices" in the oil industry. Reviewing the literature of the offshore drilling industry indicates that most of the developed risk analysis methodologies do not fully and more importantly, systematically address the contribution of Human and Organizational Factors (HOFs) in accident causation. This is while results of a comprehensive study, from 1988 to 2005, of more than 600 well-documented major failures in offshore structures show that approximately 80% of those failures were due to HOFs. In addition, lack of safety culture, as an issue related to HOFs, have been identified as a common contributing cause of many accidents in this industry. This dissertation introduces an integrated risk analysis methodology to systematically assess the critical role of human and organizational factors in offshore drilling safety. The proposed methodology in this research focuses on a specific procedure called Negative Pressure Test (NPT), as the primary method to ascertain well integrity during offshore drilling, and analyzes the contributing causes of misinterpreting such a critical test. In addition, the case study of the BP Deepwater Horizon accident and their conducted NPT is discussed. The risk analysis methodology in this dissertation consists of three different approaches and their integration constitutes the big picture of my whole methodology. The first approach is the comparative analysis of a "standard" NPT, which is proposed by the author, with the test conducted by the DWH crew. This analysis contributes to identifying the involved discrepancies between the two test procedures. The second approach is a conceptual risk assessment framework to analyze the causal factors of the identified mismatches in the previous step, as the main contributors of negative pressure test

  20. Guidelines for pressure vessel safety assessment

    Science.gov (United States)

    Yukawa, S.

    1990-04-01

    A technical overview and information on metallic pressure containment vessels and tanks is given. The intent is to provide Occupational Safety and Health Administration (OSHA) personnel and other persons with information to assist in the evaluation of the safety of operating pressure vessels and low pressure storage tanks. The scope is limited to general industrial application vessels and tanks constructed of carbon or low alloy steels and used at temperatures between -75 and 315 C (-100 and 600 F). Information on design codes, materials, fabrication processes, inspection and testing applicable to the vessels and tanks are presented. The majority of the vessels and tanks are made to the rules and requirements of ASME Code Section VIII or API Standard 620. The causes of deterioration and damage in operation are described and methods and capabilities of detecting serious damage and cracking are discussed. Guidelines and recommendations formulated by various groups to inspect for the damages being found and to mitigate the causes and effects of the problems are presented.

  1. Pressure Safety: Advanced Live 11459

    Energy Technology Data Exchange (ETDEWEB)

    Glass, George [Los Alamos National Laboratory

    2016-03-02

    Many Los Alamos National Laboratory (LANL) operations use pressure equipment and systems. Failure to follow proper procedures when designing or operating pressure systems can result in injuries to personnel and damage to equipment and/or the environment. This manual presents an overview of the requirements and recommendations that address the safe design and operation of pressure systems at LANL.

  2. Safety system for pressure suppression

    International Nuclear Information System (INIS)

    Wood, L.E.; Ludwig, G.J.; Tulsa, O.

    1975-01-01

    The rupture disk with rated breaking points is constrained by two supporting elements and has a convex-concave shape. For pressure suppression, it is reversable inversely to its bulging. Its surface has notches which are the rated breaking points and respond to higher pressures. The centre of the rupture disk contains an area of relatively smaller thickness that will burst at lower pressure and thus makes it applicable for lower pressures. For the response of the rupture disk centre, a thrust ring with a central opening may also be used. Its edge is formed into a convex-concave section supported on the edge of the rupture disk on the exit side. The free centre of the rupture disk is then the area of relative weakness. (RW/AK) [de

  3. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  4. Safety tests file

    International Nuclear Information System (INIS)

    2011-01-01

    The design and operation of nuclear power plants is governed by strict and clearly defined regulations designed to ensure their safety in all circumstances. Since the first nuclear reactors were commissioned, the basic safety principles and the corresponding practical requirements have constantly evolved and been enhanced, benefiting from operating experience feedback from reactors around the world (about 500 production reactors currently in service). Reactor safety has from the outset been built around the 'defense in depth' concept, which aims to prevent melting of the core and radioactive releases into the environment. It can be summarized as follows: over and above all the measures taken to prevent accidents, the principle that accidents do occur has to be accepted. We then assess their consequences and take steps to contain them at the level of severity at which they occur. (authors)

  5. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  6. 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

  7. Safety of pulmonary function testing

    DEFF Research Database (Denmark)

    Roberts, Cara; Ward, Simon; Walsted, Emil

    2017-01-01

    BACKGROUND: Pulmonary function testing (PFT) is a key investigation in the evaluation of individuals with respiratory symptoms; however, the safety of routine and specialised PFT testing has not been reported in a large data set. Using patient safety incident (PSI) records, we aimed to assess risk...... was rated using the NHS National Patient Safety Agency and any hospital admission reported. RESULTS: There were 119 PSIs reported from 186 000 PFT; that is, 0.6 PSIs per 1000 tests. Cardiopulmonary PSIs were 3.3 times more likely to occur than non-cardiopulmonary (95% CI 2.17 to 5.12). Syncope was the most...

  8. Development of main steam safety valve set pressure evaluating system

    International Nuclear Information System (INIS)

    Oketani, Koichiro; Manabe, Yoshihisa.

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

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

  10. Pressure Safety: Advanced Self-Study 30120

    Energy Technology Data Exchange (ETDEWEB)

    Glass, George [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-02-29

    Pressure Safety Advance Self-Study (Course 30120) consists of an introduction, five modules, and a quiz. To receive credit in UTrain for completing this course, you must score 80% or better on the 15-question quiz (check UTrain). Directions for initiating the quiz are appended to the end of this training manual. This course contains several links to LANL websites. UTrain might not support active links, so please copy links into the address line in your browser.

  11. Pressure test method for reactor pressure vessel in construction field

    International Nuclear Information System (INIS)

    Takeda, Masakado; Ushiroda, Koichi; Miyahara, Ryohei; Takano, Hiroshi; Matsuura, Tadashi; Sato, Keiya.

    1998-01-01

    Plant constitutional parts as targets of both of a primary pressure test and a secondary pressure test are disposed in communication with a reactor pressure vessel, and a pressure of the primary pressure test is applied to the targets of both tests, so that the primary pressure test and the second pressure test are conducted together. Since the number of pressure tests can be reduced to promote construction, and the number of workers can also be reduced. A pressure exceeding the maximum pressure upon use is applied to the pressure vessel after disposing the incore structures, to continuously conduct the primary pressure test and the secondary pressure test joined together and an incore flowing test while closing the upper lid of the pressure vessel as it is in the construction field. The number of opening/closing of the upper lid upon conducting every test can be reduced, and since the pressure resistance test is conducted after arranging circumference conditions for the incore flowing test, the tests can be conducted collectively also in view of time. (N.H.)

  12. Analysis on working pressure selection of ACME integral test facility

    International Nuclear Information System (INIS)

    Chen Lian; Chang Huajian; Li Yuquan; Ye Zishen; Qin Benke

    2011-01-01

    An integral effects test facility, advanced core cooling mechanism experiment facility (ACME) was designed to verify the performance of the passive safety system and validate its safety analysis codes of a pressurized water reactor power plant. Three test facilities for AP1000 design were introduced and review was given. The problems resulted from the different working pressures of its test facilities were analyzed. Then a detailed description was presented on the working pressure selection of ACME facility as well as its characteristics. And the approach of establishing desired testing initial condition was discussed. The selected 9.3 MPa working pressure covered almost all important passive safety system enables the ACME to simulate the LOCAs with the same pressure and property similitude as the prototype. It's expected that the ACME design would be an advanced core cooling integral test facility design. (authors)

  13. Cadmium safety rod thermal tests

    International Nuclear Information System (INIS)

    Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

    1992-01-01

    Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

  14. High temperature and pressure electrochemical test station

    DEFF Research Database (Denmark)

    Chatzichristodoulou, Christodoulos; Allebrod, Frank; Mogensen, Mogens Bjerg

    2013-01-01

    An electrochemical test station capable of operating at pressures up to 100 bars and temperatures up to 400 ◦C has been established. It enables control of the partial pressures and mass flow of O2, N2, H2, CO2, and H2O in a single or dual environment arrangement, measurements with highly corrosive...... media, as well as localized sampling of gas evolved at the electrodes for gas analysis. A number of safety and engineering design challenges have been addressed. Furthermore, we present a series of electrochemical cell holders that have been constructed in order to accommodate different types of cells...... and facilitate different types of electrochemical measurements. Selected examples of materials and electrochemical cells examined in the test station are provided, ranging from the evaluation of the ionic conductivity of liquid electrolytic solutions immobilized in mesoporous ceramic structures...

  15. High-pressure safety at the Lawrence Livermore Laboratory, an energy research facility

    International Nuclear Information System (INIS)

    Burton, W.A.

    1976-01-01

    The high-pressure safety program at Lawrence Livermore Laboratory, Livermore, California, has been successful in preventing lost-time high-pressure accidents over the past 12 years. Program organization, personnel training and qualification, pressure vessel design criteria and documentation, and pressure testing and inspection are discussed

  16. Triplets pass their pressure test

    CERN Multimedia

    2007-01-01

    All the LHC inner triplets have now been repaired and are in position. The first ones have passed their pressure tests with flying colours. The repaired inner triplet at LHC Point 1, right side (1R). Ranko Ostojic (on the right), who headed the team responsible for repairing the triplets, shows the magnet to Robert Zimmer, President of the University of Chicago and of Fermi Research Alliance, who visited CERN on 20th August.Three cheers for the triplets! All the LHC inner triplets have now been repaired and are in position in the tunnel. Thanks to the mobilisation of a multidisciplinary team from CERN and Fermilab, assisted by the KEK Laboratory and the Lawrence Berkeley National Laboratory (LBNL), a solution has been found, tested, validated and applied. At the end of March this year, one of the inner triplets at Point 5 failed to withstand a pressure test. A fault was identified in the supports of two out of the three quadruple magne...

  17. Systematic impact of institutional pressures on safety climate in the construction industry.

    Science.gov (United States)

    He, Qinghua; Dong, Shuang; Rose, Timothy; Li, Heng; Yin, Qin; Cao, Dongping

    2016-08-01

    This paper explores how three types of institutional pressure (i.e., coercive, mimetic and normative pressures) systematically impact on the safety climate of construction projects. These impacts are empirically tested by survey data collected from 186 questionnaires of construction companies operating in Shanghai, China. The results, obtained by partial least squares analysis, show that organizational management commitment to safety and employee involvement is positively related to all three institutional pressures, while the perception of responsibility for safety and health is significantly influenced by coercive and mimetic pressure. However, coercive and normative pressures have no significant effect on the applicability of safety rules and work practices, revealing the importance of external organizational pressures in improving project safety climate from a systematic view. The findings also provide insights into the use of institutional forces to facilitate the improvement of safety climate in the construction industry. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Safety supervision on high-pressure gas regulations

    International Nuclear Information System (INIS)

    Lee, Won Il

    1991-01-01

    The first part lists the regulation on safety supervision of high-pressure gas, enforcement ordinance on high-pressure gas safety supervision and enforcement regulations about high-pressure gas safety supervision. The second part indicates safety regulations on liquefied petroleum gas and business, enforcement ordinance of safety on liquefied petroleum gas and business, enforcement regulation of safety supervision over liquefied petroleum gas and business. The third part lists regulation on gas business, enforcement ordinance and enforcement regulations on gas business. Each part has theory and explanation for questions.

  19. Safety regulation on high-pressure gas and gas business

    International Nuclear Information System (INIS)

    Kim, Du Yeoung; An, Dae Jun

    1978-09-01

    This book is divided into two parts. The first part introduces safety regulation on high-pressure gas, enforcement ordinance on safety regulation about high-pressure gas and enforcement regulation on safety regulation about high-pressure gas. The second part indicates regulations on gas business such as general rules, gas business gas supplies, using land, supervision, supple mentary rules and penalty. It has two appendixes on expected questions and questions during last years.

  20. 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

  1. Enhancement of pressurizer safety valve operability by seating design improvement

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Safety shield for vacuum/pressure-chamber windows

    Science.gov (United States)

    Shimansky, R. A.; Spencer, R.

    1980-01-01

    Optically-clear shatter-resistant safety shield protects workers from implosion and explosion of vacuum and pressure windows. Plastic shield is inexpensive and may be added to vacuum chambers, pressure chambers, and gas-filling systems.

  3. Pressure-Application Device for Testing Pressure Sensors

    Science.gov (United States)

    2002-01-01

    A portable pressure-application device has been designed and built for use in testing and calibrating piezoelectric pressure transducers in the field. The device generates pressure pulses of known amplitude. A pressure pulse (in contradistinction to a steady pressure) is needed because in the presence of a steady pressure, the electrical output of a piezoelectric pressure transducer decays rapidly with time. The device includes a stainless- steel compressed-air-storage cylinder of 500 cu cm volume. A manual hand pump with check valves and a pressure gauge are located at one end of the cylinder. A three-way solenoid valve that controls the release of pressurized air is located at the other end of the cylinder. Power for the device is provided by a 3.7-V cordless-telephone battery. The valve is controlled by means of a pushbutton switch, which activates a 5 V to +/-15 V DC-to-DC converter that powers the solenoid. The outlet of the solenoid valve is connected to the pressure transducer to be tested. Before the solenoid is energized, the transducer to be tested is at atmospheric pressure. When the solenoid is actuated by the push button, pressurized air from inside the cylinder is applied to the transducer. Once the pushbutton is released, the cylinder pressure is removed from the transducer and the pressurized air applied to the transducer is vented, bringing the transducer back to atmospheric pressure. Before this device was used for actual calibration, its accuracy was checked with a NIST (National Institute of Standards and Technology) traceable calibrator and commercially calibrated pressure transducers. This work was done by Wanda Solano of Stennis Space Center and Greg Richardson of Lockheed Martin Corp.

  4. Large-Scale Spacecraft Fire Safety Tests

    Science.gov (United States)

    Urban, David; Ruff, Gary A.; Ferkul, Paul V.; Olson, Sandra; Fernandez-Pello, A. Carlos; T'ien, James S.; Torero, Jose L.; Cowlard, Adam J.; Rouvreau, Sebastien; Minster, Olivier; hide

    2014-01-01

    An international collaborative program is underway to address open issues in spacecraft fire safety. Because of limited access to long-term low-gravity conditions and the small volume generally allotted for these experiments, there have been relatively few experiments that directly study spacecraft fire safety under low-gravity conditions. Furthermore, none of these experiments have studied sample sizes and environment conditions typical of those expected in a spacecraft fire. The major constraint has been the size of the sample, with prior experiments limited to samples of the order of 10 cm in length and width or smaller. This lack of experimental data forces spacecraft designers to base their designs and safety precautions on 1-g understanding of flame spread, fire detection, and suppression. However, low-gravity combustion research has demonstrated substantial differences in flame behavior in low-gravity. This, combined with the differences caused by the confined spacecraft environment, necessitates practical scale spacecraft fire safety research to mitigate risks for future space missions. To address this issue, a large-scale spacecraft fire experiment is under development by NASA and an international team of investigators. This poster presents the objectives, status, and concept of this collaborative international project (Saffire). The project plan is to conduct fire safety experiments on three sequential flights of an unmanned ISS re-supply spacecraft (the Orbital Cygnus vehicle) after they have completed their delivery of cargo to the ISS and have begun their return journeys to earth. On two flights (Saffire-1 and Saffire-3), the experiment will consist of a flame spread test involving a meter-scale sample ignited in the pressurized volume of the spacecraft and allowed to burn to completion while measurements are made. On one of the flights (Saffire-2), 9 smaller (5 x 30 cm) samples will be tested to evaluate NASAs material flammability screening tests

  5. Space station pressurized laboratory safety guidelines

    Science.gov (United States)

    Mcgonigal, Les

    1990-01-01

    Before technical safety guidelines and requirements are established, a common understanding of their origin and importance must be shared between Space Station Program Management, the User Community, and the Safety organizations involved. Safety guidelines and requirements are driven by the nature of the experiments, and the degree of crew interaction. Hazard identification; development of technical safety requirements; operating procedures and constraints; provision of training and education; conduct of reviews and evaluations; and emergency preplanning are briefly discussed.

  6. Enhancement of pressurizer safety valve operability by seating design improvement

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Safety in the use of pressurized suits

    International Nuclear Information System (INIS)

    1984-01-01

    This Code of Practice describes the procedures relating to the safe operation of Pressurized Suit Areas and their supporting services. It is directed at personnel responsible for the design and/or operation of Pressurized Suit Areas. (author)

  8. Full-scale aircraft tire pressure tests

    OpenAIRE

    FABRE, C; BALAY, Jean Maurice; LERAT, P; MAZARS, A

    2009-01-01

    This paper describes an outdoor full-scale test planned to improve experimental and theoretical knowledge related to the effects of aircraft internal tire inflation pressure on the behavior and damage of flexible pavement. Since modern aircraft can have tire pressures greater than 15 bar, the tests will focus on pressures from 15 bar to 17.5 bar. The experimental pavement located on the Toulouse-Blagnac airport in France will include up to seven al different test sections, representative of c...

  9. Full-scale aicraft tire pressure tests

    OpenAIRE

    FABRE, C; BALAY, Jean Maurice; LERAT, P; MAZARS, A

    2009-01-01

    This paper describes an outdoor full-scale test planned to improve experimental and theoretical knowledge related to the effects of aircraft internal tire inflation pressure on the behavior and damage of flexible pavement. Since modern aircraft can have tire pressures greater than 15 bar, the tests focus on pressures from 15 to 17.5 bar. The experimental pavement located on the Toulouse-Blagnac airport in France will include up to seven al different test sections, representative of current ai...

  10. The pressure and leak tests in Atucha II

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    This work deals with the pressure and leak tests of the containment sphere in the Atucha II Nuclear Power Plant's reactor building. This sphere is a metallic container, made in highly resistant steel plate, that is, built for providing the plant with a biological and structural barrier, which -in turn- provides safety and environmental protection. The applicable rules for these tests establish that the containment erection must be complete and in equivalent conditions to those that will prevail during the NPP operation. Particularly, pressure tests were carried out for assessing the structural condition of the sphere, while the leak test is aimed at the detection of tentative leaks [es

  11. Model tests for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Stoever, R.

    1975-01-01

    Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

  12. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Forgie, J.H.; Holland, J.A.

    1983-01-01

    In the period when PCRV were relatively unproven, proof pressure testing provided a useful demonstration of vessel integritiy and a confirmation of model testing and of analysis. No failures have occurred during concrete vessel tests in the UK or in the subsequent operational life of the vessels and much has been learned of their behaviour in service. The paper examines the advantages and disadvantages of proof testing PCRV in the light of the above increased knowledge of vessel performance. The paper draws attention to certain hypothetical loading cases that could be more onerous than the proof test and suggests that pressure testing could itself cause unnecessarily high loading to parts of the vessel. Always recognising the safety considerations and demonstrations of such are of prime importance, the authors suggest that a lower pressure level could be adopted without loss of original intent. In addition some ground rules are suggested as to cases where proof testing could be omitted. (orig./HP)

  13. Radiation pressure actuation of test masses

    International Nuclear Information System (INIS)

    Garoi, F; Ju, L; Zhao, C; Blair, D G

    2004-01-01

    In this paper, we investigate the use of radiation pressure force as test mass actuation for laser interferometer gravitational wave detectors. It is shown that it is viable to provide radiation pressure control on test masses for frequencies above ∼0.2 Hz in high performance vibration isolation systems. A very low mass, low frequency resonator has been used to verify that radiation pressure force is not corrupted by other forces such as due to radiometer effects

  14. Safety in acoustic emission testing

    International Nuclear Information System (INIS)

    Pollock, A.A.

    2004-01-01

    The human cost of accidents - the loss of family members and friends, and the impairment of life through injuries - makes prevention a very high priority in our society. Especially in the more industrialized countries, where high levels of personal comfort and security are the norm, resources are available to develop safety-enhancing technologies, cultures and management techniques. Thus, safety programs have become a well-established part of the industrial workplace. (author)

  15. Profit and safety - the pressures on the manufacturer

    Energy Technology Data Exchange (ETDEWEB)

    Hedley, D.A.; Stace, L.R. [EXM Ltd., Uttoxeter (United Kingdom)

    2002-04-01

    Aspects of the introduction of new technology into the UK coal mining industry at a time when both suppliers and mine operators face considerable financial pressures are reviewed, mainly from the viewpoint of the manufacturer. Three examples of products supplied by one manufacturer are presented to illustrate the difficulties that a supplier faces. The criteria that are applied to judge the suitability of new designs are discussed and the need for a constant review of the industry's long-standing technical specification and test requirements is highlighted. The safety culture prevailing in the UK mining industry is justifiably a source of pride, but the continuing introduction of additional Directives and technical standards by the European Commission (EC) and its agencies could place additional costs upon the manufacturer. Burgeoning administrative tasks in the areas of taxation and pensions add to the problems that are faced by small companies. Three products, all supplied by Hanning Ltd., are described as examples of the effects of external regulatory pressures on product design - the cable handling monorail, the underground multimeter and the light filter. The paper was presented to the Institution of Mining and Metallurgy's South Midlands Branch joint symposium with the Health and Safety Executive and Nottingham University, 25 March 1999 in Nottingham, UK. 6 figs.

  16. Tests Of Array Of Flush Pressure Sensors

    Science.gov (United States)

    Larson, Larry J.; Moes, Timothy R.; Siemers, Paul M., III

    1992-01-01

    Report describes tests of array of pressure sensors connected to small orifices flush with surface of 1/7-scale model of F-14 airplane in wind tunnel. Part of effort to determine whether pressure parameters consisting of various sums, differences, and ratios of measured pressures used to compute accurately free-stream values of stagnation pressure, static pressure, angle of attack, angle of sideslip, and mach number. Such arrays of sensors and associated processing circuitry integrated into advanced aircraft as parts of flight-monitoring and -controlling systems.

  17. Safety surveillance of activities on nuclear pressure components in China

    International Nuclear Information System (INIS)

    Li Ganjie; Li Tianshu; Yan Tianwen

    2005-01-01

    The nuclear pressure components, which perform the nuclear safety functions, are one of the key physical barriers for nuclear safety. For the national strategy on further development of nuclear power and localization of nuclear pressure components, there still exist some problems in preparedness on the localization. As for the technical basis, what can not be overlooked is the management. Aiming at the current problems, National Nuclear Safety Administration (NNSA) has taken measures to strengthen the propagation and popularization of nuclear safety culture, adjust the review and approval policies for nuclear pressure components qualification license, establish more stringent management requirements, and enhance the surveillance of activities on nuclear pressure equipment. Meanwhile, NNSA has improved the internal management and the regulation efficiency on nuclear pressure components. At the same time, with the development and implementation of 'Rules on the Safety Regulation for Nuclear Safety Important Components' to be promulgated by the State Council of China, NNSA will complete and improve the regulation on nuclear pressure components and other nuclear equipment. (authors)

  18. Multi-Canister overpack pressure testing

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The Multi-Canister Overpack (MCO) shield plug closure assembly will be hydrostatically tested at the fabricator's shop to the 150 psig design test requirement in accordance with the ASME Code. Additionally, the MCO shell and collar will be hydrostatically tested at the fabricator's shop to the 450 psig design test requirement. Commercial practice has not required a pressure test of the closure weld after spent fuel is loaded in the containers. Based on this precedent and Code Case N-595-I, the MCO closure weld will not be pressure tested in the field

  19. Safety and reliability of pressure components with special emphasis on the contribution of component and large specimen testing to structural integrity assessment methodology. Vol. 1 and 2

    International Nuclear Information System (INIS)

    1987-01-01

    The 51 papers of the 13. MPA-seminar contribute to structural integrity assessment methodology with special emphasis on the component and large specimen testing. 8 of the papers deal with fracture mechanics, 6 papers with dynamic loading, 13 papers with nondestructive testing, 2 papers with radiation embrittlement, 5 papers with pipe failure, 4 papers with components, 2 papers with thermal shock loading, 5 papers with the high temperature behaviour, 4 papers with the integrity of vessels and 3 papers with the integrity of welded joints. Especially also the fracture behaviour of steel material is verificated. All papers are separately indexed and analysed for the database. (DG) [de

  20. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  1. EDF'S nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1988-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

  2. Drilling and testing hot, high-pressure wells

    Energy Technology Data Exchange (ETDEWEB)

    MacAndrew, R. (Ranger Oil Ltd, Aberdeen (United Kingdom)); Parry, N. (Phillips Petroleum Company United Kingdom Ltd, Aberdeen (United Kingdom)); Prieur, J.M. (Conoco UK Ltd, Aberdeen (United Kingdom)); Wiggelman, J. (Shell UK Exploration and Production, Aberdeen (United Kingdom)); Diggins, E. (Brunei Shell Petroleum (Brunei Darussalam)); Guicheney, P. (Sedco Forex, Montrouge (France)); Cameron, D.; Stewart, A. (Dowell Schlumberger, Aberdeen (United Kingdom))

    Meticulous planning and careful control of operations are needed to safely drill and test high-temperature, high-pressure (HTHP) wells. Techniques, employed in the Central Graben in the UK sector of the North Sea, where about 50 HTHP wells have been drilled, are examined. Three main areas of activity are covered in this comprehensive review: drilling safety, casing and cementation, and testing. The three issues at the heart of HTHP drilling safety are kick prevention, kick detection and well control. Kicks are influxes of reservoir fluid into the well. Test equipment and operations are divided into three sections: downhole, subsea and surface. Also details are given of how this North Sea experience has been used to help plan a jackup rig modification for hot, high-pressure drilling off Brunei. 16 figs., 32 refs.

  3. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

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

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

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

  5. Model-based testing for software safety

    NARCIS (Netherlands)

    Gurbuz, Havva Gulay; Tekinerdogan, Bedir

    2017-01-01

    Testing safety-critical systems is crucial since a failure or malfunction may result in death or serious injuries to people, equipment, or environment. An important challenge in testing is the derivation of test cases that can identify the potential faults. Model-based testing adopts models of a

  6. Commonwealth Edison Company pressure locking test report

    Energy Technology Data Exchange (ETDEWEB)

    Bunte, B.D.; Kelly, J.F.

    1996-12-01

    Pressure Locking is a phenomena which can cause the unseating thrust for a gate valve to increase dramatically from its typical static unseating thrust. This can result in the valve actuator having insufficient capability to open the valve. In addition, this can result in valve damage in cases where the actuator capability exceeds the valve structural limits. For these reasons, a proper understanding of the conditions which may cause pressure locking and thermal binding, as well as a methodology for predicting the unseating thrust for a pressure locked or thermally bound valve, are necessary. This report discusses the primary mechanisms which cause pressure locking. These include sudden depressurization of piping adjacent to the valve and pressurization of fluid trapped in the valve bonnet due to heat transfer. This report provides a methodology for calculating the unseating thrust for a valve which is pressure locked. This report provides test data which demonstrates the accuracy of the calculation methodology.

  7. Commonwealth Edison Company pressure locking test report

    International Nuclear Information System (INIS)

    Bunte, B.D.; Kelly, J.F.

    1996-01-01

    Pressure Locking is a phenomena which can cause the unseating thrust for a gate valve to increase dramatically from its typical static unseating thrust. This can result in the valve actuator having insufficient capability to open the valve. In addition, this can result in valve damage in cases where the actuator capability exceeds the valve structural limits. For these reasons, a proper understanding of the conditions which may cause pressure locking and thermal binding, as well as a methodology for predicting the unseating thrust for a pressure locked or thermally bound valve, are necessary. This report discusses the primary mechanisms which cause pressure locking. These include sudden depressurization of piping adjacent to the valve and pressurization of fluid trapped in the valve bonnet due to heat transfer. This report provides a methodology for calculating the unseating thrust for a valve which is pressure locked. This report provides test data which demonstrates the accuracy of the calculation methodology

  8. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  9. Behaving safely under pressure: The effects of job demands, resources, and safety climate on employee physical and psychosocial safety behavior.

    Science.gov (United States)

    Bronkhorst, Babette

    2015-12-01

    Previous research has shown that employees who experience high job demands are more inclined to show unsafe behaviors in the workplace. In this paper, we examine why some employees behave safely when faced with these demands while others do not. We add to the literature by incorporating both physical and psychosocial safety climate in the job demands and resources (JD-R) model and extending it to include physical and psychosocial variants of safety behavior. Using a sample of 6230 health care employees nested within 52 organizations, we examined the relationship between job demands and (a) resources, (b) safety climate, and (c) safety behavior. We conducted multilevel analyses to test our hypotheses. Job demands (i.e., work pressure), job resources (i.e., job autonomy, supervisor support, and co-worker support) and safety climate (both physical and psychosocial safety climate) are directly associated with, respectively, lower and higher physical and psychosocial safety behavior. We also found some evidence that safety climate buffers the negative impact of job demands (i.e., work-family conflict and job insecurity) on safety behavior and strengthens the positive impact of job resources (i.e., co-worker support) on safety behavior. Regardless of whether the focus is physical or psychological safety, our results show that strengthening the safety climate within an organization can increase employees' safety behavior. Practical implication: An organization's safety climate is an optimal target of intervention to prevent and ameliorate negative physical and psychological health and safety outcomes, especially in times of uncertainty and change. Copyright © 2015 Elsevier Ltd and National Safety Council. All rights reserved.

  10. 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

  11. Justification of response time testing requirements for pressure and differential pressure sensors

    International Nuclear Information System (INIS)

    Weiss, J.M.; Mayo, C.; Swisher, V.

    1991-01-01

    This paper reports on response time testing (RTT) requirements that were imposed on pressure, differential pressure sensors as a conservative approach to insure that assumptions in the plant safety analyses were met. The purpose of this project has been to identify the need for response time testing using the bases identified in IEEE Standard 338. A combination of plant data analyses, failure modes, and effects analyses (FMEAs) was performed. Eighteen currently qualified sensor models were utilized. The results of these analyses indicate that there are only two failure modes that affect response time, not sensor output concurrently. For these failure modes, appropriate plant actions and testing techniques were identified. Safety system RTT requirements were established by IEEE Standard 338-1975. Criteria for the Periodic Testing of Class IE Power, Protection Systems, presuming the need existed for this testing. This standard established guidelines for periodic testing to verify that loop response times of installed nuclear safety-related equipment were within the limits presumed by the design basis plant transient, accident analyses. The requirements covered all passive, active components in an instrument loop, including sensors. Individual components could be tested either in groups or separately to determine the overall loop response time

  12. 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)

  13. 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)

  14. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  15. Validation testing of safety-critical software

    International Nuclear Information System (INIS)

    Kim, Hang Bae; Han, Jae Bok

    1995-01-01

    A software engineering process has been developed for the design of safety critical software for Wolsung 2/3/4 project to satisfy the requirements of the regulatory body. Among the process, this paper described the detail process of validation testing performed to ensure that the software with its hardware, developed by the design group, satisfies the requirements of the functional specification prepared by the independent functional group. To perform the tests, test facility and test software were developed and actual safety system computer was connected. Three kinds of test cases, i.e., functional test, performance test and self-check test, were programmed and run to verify each functional specifications. Test failures were feedback to the design group to revise the software and test results were analyzed and documented in the report to submit to the regulatory body. The test methodology and procedure were very efficient and satisfactory to perform the systematic and automatic test. The test results were also acceptable and successful to verify the software acts as specified in the program functional specification. This methodology can be applied to the validation of other safety-critical software. 2 figs., 2 tabs., 14 refs. (Author)

  16. High Pressure Quick Disconnect Particle Impact Tests

    Science.gov (United States)

    Rosales, Keisa R.; Stoltzfus, Joel M.

    2009-01-01

    NASA Johnson Space Center White Sands Test Facility (WSTF) performed particle impact testing to determine whether there is a particle impact ignition hazard in the quick disconnects (QDs) in the Environmental Control and Life Support System (ECLSS) on the International Space Station (ISS). Testing included standard supersonic and subsonic particle impact tests on 15-5 PH stainless steel, as well as tests performed on a QD simulator. This paper summarizes the particle impact tests completed at WSTF. Although there was an ignition in Test Series 4, it was determined the ignition was caused by the presence of a machining imperfection. The sum of all the test results indicates that there is no particle impact ignition hazard in the ISS ECLSS QDs. KEYWORDS: quick disconnect, high pressure, particle impact testing, stainless steel

  17. Safety shield for vacuum/pressure chamber viewing port

    Science.gov (United States)

    Shimansky, R. A.; Spencer, R. S. (Inventor)

    1981-01-01

    Observers are protected from flying debris resulting from a failure of a vacuum or pressure chamber viewing port following an implosion or explosion by an optically clear shatter resistant safety shield which spaced apart from the viewing port on the outer surface of the chamber.

  18. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  19. Ensuring the nuclear safety of VVER-440 reactor pressure vessels in Skoda, Concern Enterprise, Plzen

    International Nuclear Information System (INIS)

    Hrbek, Z.

    1985-01-01

    Various types of routine inspections are described of reactor pressure vessels with the aim of identifying residual lifetime and overall safety. The inspection programme includes: choice of systems and instruments, type of tests, test frequency, safety criteria, measures to be taken in case of unsatisfactory results, documentation. The criteria are given for periodical inspections and requirements listed for instruments and equipment. The main three groups of tests are: visual inspection and dimension tests, surface inspection and volumetric inspection. Briefly described is some of the equipment used. (M.D.)

  20. 49 CFR 178.814 - Hydrostatic pressure test.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Hydrostatic pressure test. 178.814 Section 178.814... Testing of IBCs § 178.814 Hydrostatic pressure test. (a) General. The hydrostatic pressure test must be... preparation for the hydrostatic pressure test. For metal IBCs, the test must be carried out before the fitting...

  1. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  2. Preservation of FFTF Data Related to Passive Safety Testing

    International Nuclear Information System (INIS)

    Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

    2010-01-01

    One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980's. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: (1) to verify natural circulation as a reliable means to safely remove decay heat, (2) to extend passive safety

  3. Simplified pressure method for respirator fit testing.

    Science.gov (United States)

    Han, D; Xu, M; Foo, S; Pilacinski, W; Willeke, K

    1991-08-01

    A simplified pressure method has been developed for fit testing air-purifying respirators. In this method, the air-purifying cartridges are replaced by a pressure-sensing attachment and a valve. While wearers hold their breath, a small pump extracts air from the respirator cavity until a steady-state pressure is reached in 1 to 2 sec. The flow rate through the face seal leak is a unique function of this pressure, which is determined once for all respirators, regardless of the respirator's cavity volume or deformation because of pliability. The contaminant concentration inside the respirator depends on the degree of dilution by the flow through the cartridges. The cartridge flow varies among different brands and is measured once for each brand. The ratio of cartridge to leakflow is a measure of fit. This flow ratio has been measured on human subjects and has been compared to fit factors determined on the same subjects by means of photometric and particle count tests. The aerosol tests gave higher values of fit.

  4. Initiation and arrest - two approaches to pressure vessel safety

    International Nuclear Information System (INIS)

    Brumovsky, M.; Filip, R.; Stepanek, S.

    1976-01-01

    The safety analysis is described of the reactor pressure vessel related to brittle fracture based on the fracture mechanics theory using two different approximations, i.e., the Crack Arrest Temperature (CAT) or Nil Ductility Temperature (NDT), and fracture toughness. The variation of CAT with stress was determined for different steel specimens of 120 to 200 mm in thickness. A diagram is shown of CAT variation with stress allowing the determination of crack arrest temperature for all types of commonly used steels independently of the NDT initial value. The diagram also shows that the difference between fracture transition elastic (FTE) and NDT depends on the type of material and determines the value of the ΔTsub(sigma) factor typical of the safety coefficient. The so-called fracture toughness reference value Ksub(IR) is recommended for the computation of pressure vessel criticality. Also shown is a defect analysis diagram which may be used for the calculation of pressure vessel safety prior to and during operation and which may also be used in making the decision on what crack sizes are critical, what cracks may be arrested and what cracks are likely to expand. The diagram is also important for the fact that it is material-independent and may be employed for the estimates of pre-operational and operational inspections and for pressure vessel life prediction. It is generally applicable to materials of greater thickness in the region where the validity of linear elastic fracture mechanics is guaranteed. (J.P.)

  5. Safety of pressurized water reactors: problems and corresponding studies

    International Nuclear Information System (INIS)

    Cogne, F.

    1976-01-01

    The author recalls the safety problems subject to researches in the CEA, either because of their importance or because studies made abroad were not sufficiently developed or were classified or in order to acquire an independent judgement when safety is concerned. Those problems and studies are submitted referring to the 3 existing shields between man and dangerous materials: fuel element can, thermal shield, (pressure vessel and pipes), biological shield of which the behaviour is studied in connection with outside aggressions such as earthquakes, plane crashes, chemical explosions.. [fr

  6. Eddy current testing of composite pressure vessels

    Science.gov (United States)

    Casperson, R.; Pohl, R.; Munzke, D.; Becker, B.; Pelkner, M.

    2018-04-01

    The use of composite pressure vessels instead of conventional vessels made of steel or aluminum grew strongly over the last decade. The reason for this trend is the tremendous weight saving in the case of composite vessels. However, the long-time behavior is not fully understood for filling and discharging cycles and creep strength and their influence on the CFRP coating (carbon fiber reinforced plastics) and the internal liner (steel, aluminum, or plastics). The CFRP ensures the pressure resistance while the inner liner is used as a container for liquid or gas. To overcome the missing knowledge of aging, BAM started an internal project to investigate degradation of these material systems. Therefore, applicable testing methods like eddy current testing are needed. Normally, high-frequency eddy current testing (HF-ET, f > 10 MHz) is deployed for CFRP due to its low conductivity of the fiber, which is in the order of 0.01 MS/s, and the capacitive coupling between the fibers. Nevertheless, in some cases conventional ET can be applied. We show a concise summary of studies on the application of conventional ET of composite pressure vessels.

  7. Energy Systems High-Pressure Test Laboratory | Energy Systems Integration

    Science.gov (United States)

    Facility | NREL Energy Systems High-Pressure Test Laboratory Energy Systems High-Pressure Test Laboratory In the Energy Systems Integration Facility's High-Pressure Test Laboratory, researchers can safely test high-pressure hydrogen components. Photo of researchers running an experiment with a hydrogen fuel

  8. 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.

  9. Apparatus for Leak Testing Pressurized Hoses

    Science.gov (United States)

    Underwood, Steve D. (Inventor); Garrison, Steve G. (Inventor); Gant, Bobby D. (Inventor); Palmer, John R. (Inventor)

    2015-01-01

    A hose-attaching apparatus for leak-testing a pressurized hose may include a hose-attaching member. A bore may extend through the hose-attaching member. An internal annular cavity may extend coaxially around the bore. At least one of a detector probe hole and a detector probe may be connected to the internal annular cavity. At least a portion of the bore may have a diameter which is at least one of substantially equal to and less than a diameter of a hose to be leak-tested.

  10. Pneumatic pressure wave generator provides economical, simple testing of pressure transducers

    Science.gov (United States)

    Gaal, A. E.; Weldon, T. P.

    1967-01-01

    Testing device utilizes the change in pressure about a bias or reference pressure level produced by displacement of a center-driven piston in a closed cylinder. Closely controlled pneumatic pressure waves allow testing under dynamic conditions.

  11. 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...

  12. Recent metal fuel safety tests in TREAT

    International Nuclear Information System (INIS)

    Wright, A.E.; Bauer, T.H.; Lo, R.K.; Robinson, W.R.; Palm, R.G.

    1986-01-01

    In-reactor safety tests have been performed on metal-alloy reactor fuel to study its response to transient-overpower conditions, in particular, the margin to cladding breach and the axial self-extrusion of fuel within intact cladding. Uranium-fissium EBR-II driver fuel elements of several burnups were tested, some to cladding breach and others to incipient breach. Transient fuel motions were monitored, and time and location of breach were measured. The test results and computations of fuel extrusion and cladding failure in metal-alloy fuel are described

  13. Logging-while-drilling (LWD) pressure test

    Energy Technology Data Exchange (ETDEWEB)

    Thirud, Aase P.

    2003-07-01

    Statoil and Halliburton have completed a successful test of a new ground-breaking formation evaluation technology on the Norwegian shelf. An LWD formation tester, the GeoTapTM sensor, was used to quantify formation pressure during drilling operations. The inaugural job was completed by Halliburton's Sperry-Sun product service line onboard the Bideford Dolphin at the Borg Field while drilling a horizontal production well in the Vigdis Extension development. The GeoTap tool, part of Sperry-Sun's StellarTM MWD/LWT suite, was run in combination with a complete logging-while-drilling sensor package and the Geo-Pilot rotary steerable drilling system. Repeat formation pressures were taken and successfully transmitted to surface. This is the first time this type of technology has been successfully applied on the Norwegian shelf.

  14. Safety testing for LHC access system

    CERN Document Server

    Valentini, F; Ninin, P; Scibile, S

    2008-01-01

    In the domain of Safety Real-Time Systems the problem of testing represents always a big effort in terms of time, costs and efficiency to guarantee an adequate coverage degree. Exhaustive tests may, in fact, not be practicable for large and distributed systems. This paper describes the testing process followed during the validation of the CERN's LHC Access System [1], responsible for monitoring and preventing physical risks for the personnel accessing the underground areas. In the paper we also present a novel strategy for the testing problem, intended to drastically reduce the time for the test patterns generation and execution. In particular, we propose a methodology for blackbox testing that relies on the application of Model Checking techniques. Model Checking is a formal method from computer science, commonly adopted to prove correctness of system’s models through an automatic system’s state space exploration against some property formulas.

  15. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  16. Combustion Safety Simplified Test Protocol Field Study

    Energy Technology Data Exchange (ETDEWEB)

    Brand, L [Gas Technology Inst., Des Plaines, IL (United States); Cautley, D. [Gas Technology Inst., Des Plaines, IL (United States); Bohac, D. [Gas Technology Inst., Des Plaines, IL (United States); Francisco, P. [Gas Technology Inst., Des Plaines, IL (United States); Shen, L. [Gas Technology Inst., Des Plaines, IL (United States); Gloss, S. [Gas Technology Inst., Des Plaines, IL (United States)

    2015-11-05

    "9Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives. A survey of state weatherization agencies on combustion safety issues, details of a field data collection instrumentation package, summary of data collected over seven months, data analysis and results are included. The project provides several key results. State weatherization agencies do not generally track combustion safety failures, the data from those that do suggest that there is little actual evidence that combustion safety failures due to spillage from non-dryer exhaust are common and that only a very small number of homes are subject to the failures. The project team collected field data on 11 houses in 2015. Of these homes, two houses that demonstrated prolonged and excessive spillage were also the only two with venting systems out of compliance with the National Fuel Gas Code. The remaining homes experienced spillage that only occasionally extended beyond the first minute of operation. Combustion zone depressurization, outdoor temperature, and operation of individual fans all provide statistically significant predictors of spillage.

  17. Liquid abrasive pressure pot scoping tests report

    International Nuclear Information System (INIS)

    Archibald, K.E.

    1996-01-01

    The primary initiatives of the LITCO Decontamination Development group at the Idaho Chemical Process Plant (ICPP) are the development of methods to eliminate the use of sodium bearing decontamination chemicals and minimization of the amount of secondary waste generated during decontamination activities. In July of 1994, a Commerce Business Daily (CBD) announcement was issued by the INEL to determine commercial interest in the development of an in-situ liquid abrasive grit blasting system. As a result of the CBD announcement, Klieber ampersand Schulz issued an Expression of Interest letter which stated they would be interested in testing a prototype Liquid Abrasive Pressure Pot (LAPP). LITCO's Decontamination group and Kleiber ampersand Schulz entered into a Cooperative Research and Development Agreement (CRADA) in which the Decontamination Development group tested the prototype LAPP in a non-radioactive hot cell mockup. Test results are provided

  18. The Safety Feature Test of QNX RTOS

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-05-15

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

  19. The Safety Feature Test of QNX RTOS

    International Nuclear Information System (INIS)

    Kim, Jang Yeol; Lee, Young Jun

    2010-01-01

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

  20. Combustion Safety Simplified Test Protocol Field Study

    Energy Technology Data Exchange (ETDEWEB)

    Brand, L. [Gas Technology Inst., Des Plaines, IL (United States); Cautley, D. [Gas Technology Inst., Des Plaines, IL (United States); Bohac, D. [Gas Technology Inst., Des Plaines, IL (United States); Francisco, P. [Gas Technology Inst., Des Plaines, IL (United States); Shen, L. [Gas Technology Inst., Des Plaines, IL (United States); Gloss, S. [Gas Technology Inst., Des Plaines, IL (United States)

    2015-11-01

    Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives. A survey of state weatherization agencies on combustion safety issues, details of a field data collection instrumentation package, summary of data collected over seven months, data analysis and results are included. The project team collected field data on 11 houses in 2015.

  1. IR1 flow tube and In-Pile Test Section Pressure drop test for the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. H.; Park, K. N.; Chi, D. Y.; Sim, B. S.; Park, S. K.; Lee, J. M.; Lee, C. Y.; Kim, H. N

    2006-02-15

    The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vertical hole call IR1 of HANARO reactor core. In order to verify the pressure drop and flow rate both the inside region of IPS at the annular region between IPS and IR1 flow tube, a pressure drop was measured by varing the flow rate on both regions. The measured pressure drop in the annular region is 209kpa at 14.9kg/s which meets the limiting condition of operation of 200kpa. The measured pressure drop in side the IPS becomes 260.25kpa which is lower than the designed value of 306.65kpa. As the pressure drop is lower than the design value, it is quite conservative from the safety and operating point of view.

  2. Safety analysis of high pressure gasous fuel container punctures

    Energy Technology Data Exchange (ETDEWEB)

    Swain, M.R. [Univ. of Miami, Coral Gables, FL (United States)

    1995-09-01

    The following report is divided into two sections. The first section describes the results of ignitability tests of high pressure hydrogen and natural gas leaks. The volume of ignitable gases formed by leaking hydrogen or natural gas were measured. Leaking high pressure hydrogen produced a cone of ignitable gases with 28{degrees} included angle. Leaking high pressure methane produced a cone of ignitable gases with 20{degrees} included angle. Ignition of hydrogen produced larger overpressures than did natural gas. The largest overpressures produced by hydrogen were the same as overpressures produced by inflating a 11 inch child`s balloon until it burst.

  3. Primary break with total loss of high pressure safety injection

    International Nuclear Information System (INIS)

    Cordelle, F.; Champ, M.; Pochard, R.

    1988-10-01

    The probabilitic safety assessment of a 900 MW plant has displayed the potential importance, with regard to the risk, of intermediate primary breaks with failure of the high pressure safety injection system. The probability of such sequence is about 10 -6 /plant X year. Therefore, it is necessary to establish: - if this sequence can lead to core melt down, - if clad ruptures can occur. This event must be taken into account to determine the repair time of contaminated systems. For these studies, a three inch equivalent diameter break is considerd, as this is the most sensitive in its category with regard to these phenomena. In addition to the above objectives, the purpose of these studies is to evaluate the sensitivity of the results to the following parameters: - the time limit at which the operator starts cooling down the plant via the steam generators. Two calculations have been made with the RELAP code (1 and 2) and two with the CATHARE code (3 and 4) - the pump trip time. Four calculations have been made with the CATHARE code (5, 6, 7 and 8). In the case of failure of only one high pressure safety injection file, 6 calculations have been made with the CATHARE code, concerning the influence of pump trip time (9, 10, 11, 12, 13 and 14)

  4. Safety Testing of Ammonium Nitrate Based Mixtures

    Science.gov (United States)

    Phillips, Jason; Lappo, Karmen; Phelan, James; Peterson, Nathan; Gilbert, Don

    2013-06-01

    Ammonium nitrate (AN)/ammonium nitrate based explosives have a lengthy documented history of use by adversaries in acts of terror. While historical research has been conducted on AN-based explosive mixtures, it has primarily focused on detonation performance while varying the oxygen balance between the oxidizer and fuel components. Similarly, historical safety data on these materials is often lacking in pertinent details such as specific fuel type, particle size parameters, oxidizer form, etc. A variety of AN-based fuel-oxidizer mixtures were tested for small-scale sensitivity in preparation for large-scale testing. Current efforts focus on maintaining a zero oxygen-balance (a stoichiometric ratio for active chemical participants) while varying factors such as charge geometry, oxidizer form, particle size, and inert diluent ratios. Small-scale safety testing was conducted on various mixtures and fuels. It was found that ESD sensitivity is significantly affected by particle size, while this is less so for impact and friction. Thermal testing is in progress to evaluate hazards that may be experienced during large-scale testing.

  5. Current state of research on pressurized water reactor safety

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

    2018-03-01

    For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

  6. System Study: High-Pressure Safety Injection 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPSI results.

  7. Industrial high pressure applications. Processes, equipment and safety

    Energy Technology Data Exchange (ETDEWEB)

    Eggers, Rudolf (ed.) [Technische Univ. Hamburg-Harburg, Hamburg (Germany). Inst. fuer Thermische Verfahrenstechnik

    2012-07-01

    Industrial high pressure processes open the door to many reactions that are not possible under 'normal' conditions. These are to be found in such different areas as polymerization, catalytic reactions, separations, oil and gas recovery, food processing, biocatalysis and more. The most famous high pressure process is the so-called Haber-Bosch process used for fertilizers and which was awarded a Nobel prize. Following an introduction on historical development, the current state, and future trends, this timely and comprehensive publication goes on to describe different industrial processes, including methanol and other catalytic syntheses, polymerization and renewable energy processes, before covering safety and equipment issues. With its excellent choice of industrial contributions, this handbook offers high quality information not found elsewhere, making it invaluable reading for a broad and interdisciplinary audience.

  8. 49 CFR 178.605 - Hydrostatic pressure test.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Hydrostatic pressure test. 178.605 Section 178.605... Testing of Non-bulk Packagings and Packages § 178.605 Hydrostatic pressure test. (a) General. The hydrostatic pressure test must be conducted for the qualification of all metal, plastic, and composite...

  9. ORION - Crew Module Side Hatch: Proof Pressure Test Anomaly Investigation

    Science.gov (United States)

    Evernden, Brent A.; Guzman, Oscar J.

    2018-01-01

    The Orion Multi-Purpose Crew Vehicle program was performing a proof pressure test on an engineering development unit (EDU) of the Orion Crew Module Side Hatch (CMSH) assembly. The purpose of the proof test was to demonstrate structural capability, with margin, at 1.5 times the maximum design pressure, before integrating the CMSH to the Orion Crew Module structural test article for subsequent pressure testing. The pressure test was performed at lower pressures of 3 psig, 10 psig and 15.75 psig with no apparent abnormal behavior or leaking. During pressurization to proof pressure of 23.32 psig, a loud 'pop' was heard at 21.3 psig. Upon review into the test cell, it was noted that the hatch had prematurely separated from the proof test fixture, thus immediately ending the test. The proof pressure test was expected be a simple verification but has since evolved into a significant joint failure investigation from both Lockheed Martin and NASA.

  10. 9 CFR 113.40 - Dog safety tests.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Dog safety tests. 113.40 Section 113... Procedures § 113.40 Dog safety tests. The safety tests provided in this section shall be conducted when... recommended for use in dogs. Serials which are not found to be satisfactory when tested pursuant to the...

  11. 46 CFR 61.40-6 - Periodic safety tests.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Periodic safety tests. 61.40-6 Section 61.40-6 Shipping... INSPECTIONS Design Verification and Periodic Testing of Vital System Automation § 61.40-6 Periodic safety tests. (a) Periodic Safety tests must demonstrate the proper operation of the primary and alternate...

  12. Research on Normal Human Plantar Pressure Test

    Directory of Open Access Journals (Sweden)

    Liu Xi Yang

    2016-01-01

    Full Text Available FSR400 pressure sensor, nRF905 wireless transceiver and MSP40 SCM are used to design the insole pressure collection system, LabVIEW is used to make HMI of data acquisition, collecting a certain amount of normal human foot pressure data, statistical analysis of pressure distribution relations about five stages of swing phase during walking, using the grid closeness degree to identify plantar pressure distribution pattern recognition, and the algorithm simulation, experimental results demonstrated this method feasible.

  13. Design and safety of the Sizewell pressurized water reactor

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

  14. Safety of light-water reactor pressure vessels against brittle fracture

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1979-01-01

    The results are surveyed of research by SKODA Trust into brittle failure resistance of materials for WWER type reactor pressure vessels and into pressure vessel operating safety. Conditions are discussed in detail decisive for initiation, propagation and arrest of brittle fracture. The tests on the Cr-Mo-V type steel showed high resistance of the steel to the formation and the propagation of brittle fracture. They also confirmed the high operating reliability and the required service life of the steel. (B.S.)

  15. Passive safety testing at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Lucoff, D.M.

    1989-01-01

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

  16. Cryogenic and Gas System Piping Pressure Tests (A Collection of PT Permits)

    International Nuclear Information System (INIS)

    Rucinski, Russell A.

    2002-01-01

    This engineering note is a collection of pipe pressure testing documents for various sections of piping for the D-Zero cryogenic and gas systems. High pressure piping must conform with FESHM chapter 5031.1. Piping lines with ratings greater than 150 psig have a pressure test done before the line is put into service. These tests require the use of pressure testing permits. It is my intent that all pressure piping over which my group has responsibility conforms to the chapter. This includes the liquid argon and liquid helium and liquid nitrogen cryogenic systems. It also includes the high pressure air system, and the high pressure gas piping of the WAMUS and MDT gas systems. This is not an all inclusive compilation of test documentation. Some piping tests have their own engineering note. Other piping section test permits are included in separate safety review documents. So if it isn't here, that doesn't mean that it wasn't tested. D-Zero has a back up air supply system to add reliability to air compressor systems. The system includes high pressure piping which requires a review per FESHM 5031.1. The core system consists of a pressurized tube trailer, supply piping into the building and a pressure reducing regulator tied into the air compressor system discharge piping. Air flows from the trailer if the air compressor discharge pressure drops below the regulator setting. The tube trailer is periodically pumped back up to approximately 2000 psig. A high pressure compressor housed in one of the exterior buildings is used for that purpose. The system was previously documented, tested and reviewed for Run I, except for the recent addition of piping to and from the high pressure compressor. The following documents are provided for review of the system: (1) Instrument air flow schematic, drg. 3740.000-ME-273995 rev. H; (2) Component list for air system; (3) Pressure testing permit for high pressure piping; (4) Documentation from Run I contained in D-Zero Engineering note

  17. 9 CFR 113.38 - Guinea pig safety test.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Guinea pig safety test. 113.38 Section... Standard Procedures § 113.38 Guinea pig safety test. The guinea pig safety test provided in this section... be injected either intramuscularly or subcutaneously into each of two guinea pigs and the animals...

  18. 9 CFR 113.33 - Mouse safety tests.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Mouse safety tests. 113.33 Section 113.33 Animals and Animal Products ANIMAL AND PLANT HEALTH INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE... Procedures § 113.33 Mouse safety tests. One of the mouse safety tests provided in this section shall be...

  19. Work Pressure and Safety Behaviors among Health Workers in Ghana: The Moderating Role of Management Commitment to Safety

    OpenAIRE

    Amponsah-Tawaih, Kwesi; Adu, Michael Appiah

    2016-01-01

    Background: safety and healthy working environment has received numerous research attention over the years. Majority of these researches seem to have been conducted in the construction industry, with little attention in the health sector. Nonetheless, there are couple of studies conducted in Africa that suggest pressure in hospitals. Therefore the aim of the study was to examine how pressure influence safety behavior in the hospitals. With reference to the relevance of safety behavior in prim...

  20. Nuclear safety: a large scale quality audit of pressurized equipment

    International Nuclear Information System (INIS)

    Faudon, Valerie

    2016-01-01

    This article notably refers to, quotes and comments a hearing organised by the French Public Office for the Assessment Scientific and Technological Choices (OPECST) on the issue of safety of pressurized equipment in nuclear reactors, and which gathered the main concerned actors (Areva, EDF, IRSN, ASN) to have an overview of quality controls in AREVA NP fabrication plants. Two different issues have been addressed: a technical metallurgical issue related to some boiler-making parts, and an issue related to quality assurance. These issues concern different older reactors (Fessenheim for example) as well as new ones (EPR Flamanville). The article indicates the different measures planned, envisaged or already implemented by the concerned actors in order to improve knowledge in the boiler-making industry, and to ensure a better quality

  1. Testing of acoustic emission method during pressure tests of WWER-440 steam generators and pressurizers

    International Nuclear Information System (INIS)

    Wuerfl, K.; Crha, J.

    1987-01-01

    The results are discussed of measuring acoustic emission in output pressure testing of steam generators and pressurizers for WWER-440 reactors. The objective of the measurements was to test the reproducibility of measurements and to find the criterion which would be used in assessing the condition of the components during manufacture and in operation. The acoustic emission was measured using a single-channel Dunegan/Endevco apparatus and a 16-channel LOCAMAT system. The results showed that after the first assembly, during a repeat dismantle of the lids and during seal replacement, processes due to seal contacts and bolt and washer deformations were the main source of acoustic emission. A procedure was defined of how to exclude new acoustic emission sources in such cases. The acoustic emission method can be used for the diagnostics of plastic deformation processes or of crack production and propagation in components during service. (Z.M.)

  2. French safety and criticality testing programmes

    International Nuclear Information System (INIS)

    Barbry, F.; Leclerc, J.; Manaranche, J.C.; Maubert, L.

    1982-01-01

    This article underlines the need to include experimental safety-criticality programmes in the French nuclear effort. The means and methods used at the Section of Experimental Nuclear Safety and Criticality Research, attached to the CEA Valduc Centre, are described. Three experimental programmes are presented: safety-criticality of the PWR fuel cycle, neutron poisoning of plutonium solutions by gadolinium and safety-criticality of slightly enriched and slightly moderated uranium oxide. Criticality accidents studies in solution are then described [fr

  3. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  4. Safeguarding the nuclear safety of WWER-440 reactor pressure vessels at SKODA Plzen

    International Nuclear Information System (INIS)

    Hrbek, Z.

    1986-01-01

    The approach is described of the SKODA enterprise to safety assurance and to providing the reliability of WWER-440 reactor pressure vessels. The philosophy is analyzed of in-service inspection and determination of the residual service life of pressure vessels. This follows up on the so-called conception of basic safety whose main aim is to preclude failures at production stage by the selection of suitable material, namely by optimizing the choice of raw materials, of metallurgical procedures such as will lead to high purity of the pressure vessel material, by introducing multiple inspection in production, reducing the sensitivity of materials to technological operations, and by high-quality welds. The quality of in-service inspections is given by the use of technical diagnostic instruments of peak quality and of modern methods of nondestructive materials testing. The instruments and methods used are described. It is stated that the experience gained with in-service inspection will make it possible to draw up operating regulations and safety criteria for nuclear installations and own inspection regulations, this with regard to technical and economic factors. (Z.M.)

  5. Thermohydraulic tests in the area of reactor safety done in CDTN

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.

    1990-01-01

    The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor) [pt

  6. 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)

  7. Safety criterion for burnout of the plate-type fuel in pressurized conditions

    International Nuclear Information System (INIS)

    Komori, Y.; Kaminaga, M.; Sakurai, F.; Ando, H.; Sudo, Y.; Saito, M.; Futamura, Y.

    1992-01-01

    The reduced enrichment program for JMTR is now underway and the core conversion to LEU (Low Enrichment Uranium) is scheduled to be made in 1993. Consistent with the safety guide which have been recently developed for research and test reactors in Japan, the safety analysis for the JMTR LEU conversion was conducted. In the safety analysis, DNB (Departure from Nucleate Boiling) heat flux correlation for the JMTR downflow condition was reconsidered because recent studies on burnout show that DNB heat fluxes with thin rectangular channels under low flow rate and low pressure conditions are much lower than predicted values by conventional DNB correlations. Available DNB data, however, are very limited for the JMTR operation pressure range, so that DNB experiments were conducted simulating the JMTR fuel subchannel. Based mainly on the present experimental data, the DNB correlations scheme composed of three correlations was selected for the JMTR safety analysis. Errors of the correlations scheme with experimental data were evaluated in order to determine the allowable limit of the minimum DNB ratio for preventing fuel failure. (author)

  8. Laboratory test requesting appropriateness and patient safety

    CERN Document Server

    Blasco, Álvaro; Carratalá, Arturo; Lopez-Garrígos, Maite; Rodriguez-Borja, Enrique

    2016-01-01

    Patient Safety emphasizes the reporting, analysis and prevention of medical errors that very often leads to adverse healthcare situations.1 in 10 patients are impacted by medical errors.The WHO calls the patient safety issue an endemic concern. A number of well-known experts of all areas in the medical field have collectedvery valuable information for a better patient treatment and higher safety culture in all medical disciplines.

  9. Radiation safety in welding and testing

    International Nuclear Information System (INIS)

    King, B.E.; Malaxos, M.; Hartley, B.M.

    1985-01-01

    There are a number of ways of achieving radiation safety in the workplace. The first is by engineering radiation safety into the equipment, providing shielded rooms and safety interlocks. The second is by following safe working procedures. The National Health and Medical Research Council's Code of practice for the control and safe handling of sealed radioactive sources used in industrial radiography (1968) sets out the standards which must be met by equipment to be used in industrial radiography

  10. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  11. Bursting tests on pressure vessels with cracks differing in configuration and location

    International Nuclear Information System (INIS)

    Stahlberg, R.

    1978-01-01

    For assessing the safety of nuclear pressure vessels exhibiting cracks, bursting test were carried out on a series of medium-size pressure vessels with and without welded nozzles and exhibiting cracks differing in configuration and location. The linear-elastic approach proved to be sufficiently accurate for straight strain conditions up to the onset of general yielding. Other analytical methods were successfully used to cover the plastic region. (orig.) [de

  12. A modified isometric test to evaluate blood pressure control with ...

    African Journals Online (AJOL)

    lifting and supporting weights) and have an important influence on blood pressure, it is essential to evaluate blood pressure response to iso- metric effort. This test can reveal high blood pressure that might otherwise not be detected. Only a few ...

  13. 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

  14. Pre-service Acoustic Emission Testing for Metal Pressure Vessel

    International Nuclear Information System (INIS)

    Lee, Jong O; Yoon, Woon Ha; Lee, Tae Hee; Lee, Jong Kyu

    2003-01-01

    The field application of acoustic emission(AE) testing for brand-new metal pressure vessel were performed. We will introduce the test procedure for acoustic emission test such as instrument check distance between sensors, sensor location, whole system calibration, pressurization sequence, noise reduction and evaluation. The data of acoustic emission test contain many noise signal, these noise can be reduced by time filtering which based on the description of observation during AE test

  15. Performance Testing Methodology for Safety-Critical Programmable Logic Controller

    International Nuclear Information System (INIS)

    Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

    2009-01-01

    The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC

  16. Test facility of the VVER-440 condensation-type pressure suppression system

    International Nuclear Information System (INIS)

    Wolff, H.; Arndt, S.

    2004-01-01

    Since the early nineties, GRS has supported regulatory authorities in Central and Eastern Europe in performing safety assessments of nuclear power plants. Especially studies of the condensation-type pressure suppression system of VVER-440/V-213-type plants have been important in this respect. Major steps in demonstrating complete functioning of the condensation-type pressure suppression system under accident conditions by experiments run in the Russian large scale test facility, BC V-213, have been completed in the past two years within the framework of various international experimental programs. The test results were used to validate specifically for power plants with VVER-400/V-213 reactors the COCOSYS GRS computer code, which is used in the safety assessments. The results of recalculations of the C02 EREC test, which simulates a break of a main steam pipe, demonstrate the present state of validation of COCOSYS for VVER condensation-type pressure suppression systems. (orig.) [de

  17. Reactor pressure vessels safety and reliability - certainty and uncertainty

    International Nuclear Information System (INIS)

    O'Neil, R.

    1977-01-01

    In the paper, it is suggested that the hazard to the population which would result from vessel failure rate of the order of 10 -6 to 10 -7 per vessel year could be acceptable to society on the basis of other natural and man-made risks. The paper considers the problems of demonstrating safety by calculation based on fracture mechanics, and indicates some of the uncertainties, and inconsistencies in the theory, particularly the effect of cracks in locally degraded volumes of material. The phenomenon of crack arrest is considered, and attention is drawn to the uncertainties as indicated at least by some tests. There is need for speedy resolution of this problem. The uncertainties in material properties, heat treatment and residual stresses are considered, and a proposed upper limit for residual defects ('original sin') is proposed. (orig.) [de

  18. Allowed outage time for test and maintenance - Optimization of safety

    International Nuclear Information System (INIS)

    Cepin, M.; Mavko, B.

    1997-01-01

    The main objective of the project is the development and application of methodologies for improvement and optimization of test and maintenance activities for safety related equipment in NPPs on basis of their enhanced safety. The probabilistic safety assessment serves as a base, which does not mean the replacement of the deterministic analyses but the consideration of probabilistic safety assessment results as complement to deterministic results. 15 refs, 2 figs

  19. The Assembly and Test of Pressure Vessel for Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kennedy, Timothy C. [Oregon State University, Corvallis (United States)

    2009-02-15

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

  20. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

    2009-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

  1. Flight testing of a luminescent surface pressure sensor

    Science.gov (United States)

    Mclachlan, B. G.; Bell, J. H.; Espina, J.; Gallery, J.; Gouterman, M.; Demandante, C. G. N.; Bjarke, L.

    1992-01-01

    NASA ARC has conducted flight tests of a new type of aerodynamic pressure sensor based on a luminescent surface coating. Flights were conducted at the NASA ARC-Dryden Flight Research Facility. The luminescent pressure sensor is based on a surface coating which, when illuminated with ultraviolet light, emits visible light with an intensity dependent on the local air pressure on the surface. This technique makes it possible to obtain pressure data over the entire surface of an aircraft, as opposed to conventional instrumentation, which can only make measurements at pre-selected points. The objective of the flight tests was to evaluate the effectiveness and practicality of a luminescent pressure sensor in the actual flight environment. A luminescent pressure sensor was installed on a fin, the Flight Test Fixture (FTF), that is attached to the underside of an F-104 aircraft. The response of one particular surface coating was evaluated at low supersonic Mach numbers (M = 1.0-1.6) in order to provide an initial estimate of the sensor's capabilities. This memo describes the test approach, the techniques used, and the pressure sensor's behavior under flight conditions. A direct comparison between data provided by the luminescent pressure sensor and that produced by conventional pressure instrumentation shows that the luminescent sensor can provide quantitative data under flight conditions. However, the test results also show that the sensor has a number of limitations which must be addressed if this technique is to prove useful in the flight environment.

  2. Multiple-vent programme to test the pressure suppression system

    International Nuclear Information System (INIS)

    Aust, E.; Schwan, H.; Vollbrandt, I.

    1979-01-01

    Three pre-tests with a multiple vent configuration have been performed at the GKSS pressure suppression test facility. First test results indicate significant chugging events with occur periodically with 0.4 to 0.2 Hz. These events appear simultaneously in less than 10 ms at the exit of the three vent pipes and cause pressure pulses in the range of 3 bar. This report gives a short description of the test facility and presents the boundary conditions of the test facility and presents the boundary conditions of the three pre-tests, test results and a first valuation of the experimental informations. (orig.) [de

  3. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  4. 77 FR 2606 - Pipeline Safety: Random Drug Testing Rate

    Science.gov (United States)

    2012-01-18

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID PHMSA-2012-0004] Pipeline Safety: Random Drug Testing Rate AGENCY: Pipeline and Hazardous Materials... pipelines and operators of liquefied natural gas facilities must select and test a percentage of covered...

  5. 75 FR 9018 - Pipeline Safety: Random Drug Testing Rate

    Science.gov (United States)

    2010-02-26

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID PHMSA-2010-0034] Pipeline Safety: Random Drug Testing Rate AGENCY: Pipeline and Hazardous Materials... pipelines and operators of liquefied natural gas facilities must select and test a percentage of covered...

  6. Relevance of microbial finished product testing in food safety management

    NARCIS (Netherlands)

    Zwietering, M.H.; Jacxsens, L.; Membre, J.M.; Nauta, M.; Peterz, M.

    2016-01-01

    Management of microbiological food safety is largely based on good design of processes, products and procedures. Finished product testing may be considered as a control measure at the end of the production process. However, testing gives only very limited information on the safety status of a food.

  7. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  8. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    T. A. Tomberlin; S. B. Grover

    2004-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

  9. German boiler and pressure vessel codes and standards: materials, manufacture, testing, equipment, erection and operation

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

    The methods by which the safety objectives on the operation of steam boilers and pressure vessels in Germany can be reached are set out in Technical Rules which are compiled and established in technical committees. Typical applications are described in the Technical Rules. A chart shows how the laws, provisions and Technical Rules for the sections 'steam boiler plant' and 'pressure vessels' are interlinked. This chapter concentrates on legal aspects, materials, manufacture, testing, erection and operation of boilers and pressure vessels in Germany. (U.K.)

  10. Relevance of microbial finished product testing in food safety management

    DEFF Research Database (Denmark)

    Zwietering, Marcel H.; Jacxsens, Liesbeth; Membré, Jeanne Marie

    2016-01-01

    Management of microbiological food safety is largely based on good design of processes, products and procedures. Finished product testing may be considered as a control measure at the end of the production process. However, testing gives only very limited information on the safety status of a food......-active way by implementing an effective food safety management system. For verification activities in a food safety management system, finished product testing may however be useful. For three cases studies; canned food, chocolate and cooked ham, the relevance of testing both of finished products....... If a hazardous organism is found it means something, but absence in a limited number of samples is no guarantee of safety of a whole production batch. Finished product testing is often too little and too late. Therefore most attention should be focussed on management and control of the hazards in a more pro...

  11. Selenide isotope generator for the Galileo Mission: safety test plan

    International Nuclear Information System (INIS)

    1979-01-01

    The intent of this safety test plan is to outline particular kinds of safety tests designed to produce information which would be useful in the safety analysis process. The program deals primarily with the response of the RTG to accident environments; accordingly two criteria were established: (1) safety tests should be performed for environments which are the most critical in terms of risk contribution; and (2) tests should be formulated to determine failure conditions for critical heat source components rather than observe heat source response in reference accident environments. To satisfy criterion 1. results of a recent safety study were used to rank various accidents in terms of expected source terms. Six kinds of tests were then proposed which would provide information meeting the second criterion

  12. Performance Test Results of Safety I and C Systems of SMART MMIS

    International Nuclear Information System (INIS)

    Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong

    2011-01-01

    KAERI has developed SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, since 1997 and submitted a SSAR (Standard design Safety Analysis Report) to Korea institute of nuclear safety (KINS) at the end of 2010 for the purpose of achieving the standard design approval (SDA) by the end of 2011. SMART MMIS has been designed with fully digitalized systems. Non-safety instrumentation and control (I and C) systems are designed based on the commercial distributed control systems. The safety I and C systems are designed using a new platform that was developed and validated by KAERI. Safety I and C systems are modularized using the platform. In the protection systems (PSs), datalinks are used to transmit data in a one-way direction in order to meet the independency requirement. In the engineered safety features-component control system (ESF-CCS), network switch devices (NSDs) are used to connect the group and loop controllers. The NSD was also newly developed and validated by KAERI. After validating the platform and NSD, a test facility was developed using the platform and NSDs to validate the performance of safety I and C systems. This paper presents the development and test results from the test facility

  13. Testing of Laterally Loaded Rigid Piles with Applied Overburden Pressure

    DEFF Research Database (Denmark)

    Sørensen, Søren Peder Hyldal; Foglia, Aligi; Ibsen, Lars Bo

    2012-01-01

    Small-scale tests have been conducted for the purpose of investigating the quasi-static behaviour of laterally loaded, non-slender piles installed in cohesionless soil. For that purpose, a new and innovative test setup has been developed. The tests have been conducted in a pressure tank...... such that it was possible to apply an overburden pressure to the soil. Hereby, the traditional uncertainties related to low effective stresses for small-scale tests has been avoided. A scaling law for laterally loaded piles has been proposed based on dimensional analysis. The novel testing method has been validated against...... the test results by means of the scaling law....

  14. Testing of Laterally Loaded Rigid Piles with Applied Overburden Pressure

    DEFF Research Database (Denmark)

    Sørensen, Søren Peder Hyldal; Ibsen, Lars Bo; Foglia, Aligi

    2015-01-01

    Small-scale tests have been conducted to investigate the quasi-static behaviour of laterally loaded, non-slender piles installed in cohesionless soil. For that purpose, a new and innovative test setup has been developed. The tests have been conducted in a pressure tank such that it was possible...... to apply an overburden pressure to the soil. As a result of that, the traditional uncertainties related to low effective stresses for small-scale tests have been avoided. A normalisation criterion for laterally loaded piles has been proposed based on dimensional analysis. The test results using the novel...... testing method have been compared with the use of the normalisation criterion....

  15. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

    In Service Inspection of Reactor Pressure Vessels is mostly done with ultrasonics. Using special 2 crystal-probes good detectability is achieved for near surface defects. The problem is to detect closely spaced cracks, to decide if the defects are surface braking and, if not, to decide the remaining ligament. The purpose of this study is to investigate to what extent Eddy Current can solve these problems. Detecting surfacebreaking cracks and fields of cracks can be done using conventional Eddy Current techniques. Mapping of closely spaced cracks requires a small probe and a high frequency. Measurement of depths a larger probe, a lower frequency and knowledge of the crackfield since 2 closely spaced shallow cracks might be mistaken for one deep crack. Depths of singel cracks can be measured down to 7-8 mm. In closely spaced crackfields the depths can not be measured. The measurement is mostly based on amplitude. For not surface breaking defects the problem is to decide the ligament, i.e. the distance between surface and cracktip. To achieve good penetration a large probe, low frequency and high energy or pulsed energy is used. Ligament up to 4 mm can be measured with good accuracy. The measurements is mostly based on phase. Noise, which originates from rough surface, varied material structure and lift off, can be reduced using multi frequency mix, probe design and scanning pattern. (author)

  16. Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

    International Nuclear Information System (INIS)

    Hayakawa, Masahiko

    1981-01-01

    The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

  17. Advances made in French safety studies on pressurized water reactors

    International Nuclear Information System (INIS)

    Pelce, J.

    1979-01-01

    The programme of French safety studies on reactors is supposed to be known in its main outlines. A few recent results, obtained in different fields are presented. They concern the safety margins evaluation, the contamination transfer and the effect of external aggressions

  18. 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.)

  19. 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.)

  20. Implications of Dynamic Pressure Transducer Mounting Variations on Measurements in Pyrotechnic Test Apparatus

    Science.gov (United States)

    Dibbern, Andreas; Crisafulli, Jeffrey; Hagopia, Michael; McDougle, Stephen H.; Saulsberry, Regor L.

    2009-01-01

    Accurate dynamic pressure measurements are often difficult to make within small pyrotechnic devices, and transducer mounting difficulties can cause data anomalies that lead to erroneous conclusions. Delayed initial pressure response followed by data ringing has been observed when using miniaturized pressure transducer mounting adapters required to interface transducers to small test chambers. This delayed pressure response and ringing, combined with a high data acquisition rate, has complicated data analysis. This paper compares the output signal characteristics from different pressure transducer mounting options, where the passage distance from the transducer face to the pyrotechnic chamber is varied in length and diameter. By analyzing the data and understating the associated system dynamics, a more realistic understanding of the actual dynamic pressure variations is achieved. Three pressure transducer mounting configurations (elongated, standard, and face/flush mount) were simultaneously tested using NASA standard initiators in closed volume pressure bombs. This paper also presents results of these pressure transducer mounting configurations as a result of a larger NASA Engineering and Safety Center pyrovalve test project. Results from these tests indicate the improved performance of using face/flush mounted pressure transducers in this application. This type of mounting improved initial pressure measurement response time by approximately 19 s over standard adapter mounting, eliminating most of the lag time; provided a near step-function type initial pressure increase; and greatly reduced data ringing in high data acquisition rate systems. The paper goes on to discuss other issues associated with the firing and instrumentation that are important for the tester to understand.

  1. 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

  2. Test for radioactive material transport package safety

    International Nuclear Information System (INIS)

    Li Guoqiang; Zhao Bing; Zhang Jiangang; Wang Xuexin; Ma Anping

    2012-01-01

    Regulations on radioactive material transport in China were introduced. Test facilities and data acquiring instruments for radioactive material package in China Institute for Radiation Protection were also introduced in this paper, which were used in drop test and thermal test. Test facilities were constructed according to the requirements of IAEA's 'Regulations for the Safe Transport of Radioactive Material' (TS-R-l) and Chinese 'Regulations for the Safe Transport of Radioactive Material' (GB 11806-2004). Drop test facilities were used in free drop test, penetration test, mechanical test (free drop test Ⅰ, free drop test Ⅱ and free drop test Ⅲ) of type A and type B packages weighing less than thirteen tons. Thermal test of type B packages can be carried out in the thermal test facilities. Certification tests of type FCo70-YQ package, type 30A-HB-01 package, type SY-I package and type XAYT-I package according to regulations were done using these facilities. (authors)

  3. Safety Assessment for transient event occurred during the ASTS test of Hanbit Unit 2

    International Nuclear Information System (INIS)

    Yang, Changkeun; Kim, Yohan; Ha, Sangjun

    2014-01-01

    Safety Injection has been actuated during the ASTS (Automatic Seismic Trip System) test of Hanbit Unit 2 on Feb. 28, 2014. It could be bad effect on system integrity. KHNP has been performed safety assessment of system for effect of Safety Injection (SI) actuation occurred during the ASTS test of hanbit Unit 2. Stable state of nuclear power plant system has been confirmed according to Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the power plant is in a safe condition were conformed. After ASTS action, thermal elimination has been processed throughout the turbine until turbine signal occurrence because ASTS is connected to M-G set in the present hanbit Unit 2. Therefore, Safety Injection signal has been actuated by rapid reduction of Steam Generator pressure. In this paper, it is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation. Stable state of nuclear power plant system has been confirmed for Safety Injection and reactor trip event occurred during the ASTS test of hanbit Unit 2. In the result of system safety assessment, major variables of nuclear power plant are located in optimal range and not exceed safety limit. It remains nuclear fuel and the integrity of the plant is in a safe condition were conformed. It is concluded that consideration of equipment and setpoint is needed for that Safety Injection has been not occurred under the unnecessary situation

  4. Test process for the safety-critical embedded software

    International Nuclear Information System (INIS)

    Sung, Ahyoung; Choi, Byoungju; Lee, Jangsoo

    2004-01-01

    Digitalization of nuclear Instrumentation and Control (I and C) system requires high reliability of not only hardware but also software. Verification and Validation (V and V) process is recommended for software reliability. But a more quantitative method is necessary such as software testing. Most of software in the nuclear I and C system is safety-critical embedded software. Safety-critical embedded software is specified, verified and developed according to V and V process. Hence two types of software testing techniques are necessary for the developed code. First, code-based software testing is required to examine the developed code. Second, after code-based software testing, software testing affected by hardware is required to reveal the interaction fault that may cause unexpected results. We call the testing of hardware's influence on software, an interaction testing. In case of safety-critical embedded software, it is also important to consider the interaction between hardware and software. Even if no faults are detected when testing either hardware or software alone, combining these components may lead to unexpected results due to the interaction. In this paper, we propose a software test process that embraces test levels, test techniques, required test tasks and documents for safety-critical embedded software. We apply the proposed test process to safety-critical embedded software as a case study, and show the effectiveness of it. (author)

  5. MODEL TESTING OF LOW PRESSURE HYDRAULIC TURBINE WITH HIGHER EFFICIENCY

    Directory of Open Access Journals (Sweden)

    V. K. Nedbalsky

    2007-01-01

    Full Text Available A design of low pressure turbine has been developed and it is covered by an invention patent and a useful model patent. Testing of the hydraulic turbine model has been carried out when it was installed on a vertical shaft. The efficiency was equal to 76–78 % that exceeds efficiency of the known low pressure blade turbines. 

  6. Stress Tests Worldwide - IAEA Nuclear Safety Action Plan

    International Nuclear Information System (INIS)

    Lyons, J.E.

    2012-01-01

    The IAEA nuclear safety action plan relies on 11 important issues. 1) Safety assessments in light of the Fukushima accident: the IAEA secretariat will develop a methodology for stress tests against specific extreme natural hazards and will provide assistance for their implementation; 2) Strengthen existing IAEA peer reviews; 3) Emergency preparedness and response; 4) National Regulatory bodies in terms of independence and adequacy of human and financial resources; 5) The development of safety culture and scientific and technical capacity in Operating Organizations; 6) The upgrading of IAEA safety standards in a more efficient way; 7) A better implementation of relevant conventions concerning nuclear safety and nuclear accidents; 8) To provide a broad assistance on safety standard for countries embarking on a nuclear power program; 9) To facilitate the use of available information, expertise and techniques concerning radiation protection; 10) To enhance the transparency of nuclear industry; and 11) To promote the cooperation between member states in nuclear safety. (A.C.)

  7. Safety measures for integrity test apparatus for IS process. Sulfuric acid decomposition section

    International Nuclear Information System (INIS)

    Noguchi, Hiroki; Kubo, Shinji; Iwatsuki, Jin; Onuki, Kaoru

    2013-07-01

    Hazardous substances such as sulfuric acid, sulfur dioxide and hydrogen iodide acid are employed in thermochemical Iodine-Sulfur (IS) process. It is necessary to take safety measure against workers and external environments to study experimentally on IS process. Presently we have been conducting to verify the soundness of main components made of engineering material in actual corrosive condition. An integrity test apparatus for the components of sulfuric acid decomposition was set up. We will use the hazardous substances such as sulfuric acid and sulfur dioxide and perform the experiment in pressurized condition in this integrity test. Safety measures for the test apparatus, operation and abnormal situation were considered prior to starting the test. This report summarized the consideration results for the safety measures on the integrity test apparatus for the components of sulfuric acid decomposition. (author)

  8. Helium leak testing of large pressure vessels or subassemblies

    International Nuclear Information System (INIS)

    Hopkins, J.S.; Valania, J.J.

    1977-01-01

    Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

  9. 76 FR 1504 - Pipeline Safety: Establishing Maximum Allowable Operating Pressure or Maximum Operating Pressure...

    Science.gov (United States)

    2011-01-10

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No... Mitigation AGENCY: Pipeline and Hazardous Materials Safety Administration (PHMSA); DOT. ACTION: Notice... system. To that end, the Hazardous Liquid and Gas Transmission Pipeline Integrity Management (IM...

  10. Safety evaluation report. Fast Flux Test Facility. Project No. 448

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    Information on the safety of the FFTF Reactor is presented under the following chapter headings: site characteristics; design of structures, components, equipment, and systems; reactor; reactor coolant system and connected systems; engineered safety features; electric power; auxiliary systems; radioactive waste management systems; radiation protection; conduct of operations; initial test programs; accident analysis; and quality assurance.

  11. Safety evaluation report. Fast Flux Test Facility. Project No. 448

    International Nuclear Information System (INIS)

    1978-01-01

    Information on the safety of the FFTF Reactor is presented under the following chapter headings: site characteristics; design of structures, components, equipment, and systems; reactor; reactor coolant system and connected systems; engineered safety features; electric power; auxiliary systems; radioactive waste management systems; radiation protection; conduct of operations; initial test programs; accident analysis; and quality assurance

  12. Finite-element pre-analysis for pressurized thermoshock tests

    International Nuclear Information System (INIS)

    Keinaenen, H.; Talja, H.; Lehtonen, M.; Rintamaa, R.; Bljumin, A.; Timofeev, B.

    1992-05-01

    The behaviour of a model pressure vessel is studied in a pressurized thermal shock loading. The tests were performed at the Prometey Institute in St. Petersburg. The calculations were performed at the Technical Research Centre of Finland. The report describes the preliminary finite-element analyses for the fourth, fifth and sixth thermoshock tests with the first model pressure vessel. Seven pressurized thermoshock tests were made with the same model using five different flaw geometries. In the first three tests the flaw was actually a blunt notch. In the two following tests (tests 4 and 5) a sharp pre-crack was produced before the test. In the last two test (tests 6 and 7) the old crack was used. According to the measurements and post-test ultrasonic examination of the crack front, the sixth test led to significant crack extension. Both temperatures and stresses were calculated using the finite-element method. The calculations were made using the idealized initial flaw geometry and preliminary material data. Both two-and three-dimensional models were used in the calculations. J-integral values were calculated from the elastic-plastic finite-element results. The stress intensity factor values were evaluated on the basis of the calculated J-integrals and compared with the preliminary material fracture toughness data obtained from the Prometey Institute

  13. Pressure test behaviour of embalse nuclear power plant containment structure

    International Nuclear Information System (INIS)

    Bruschi, S.; Marinelli, C.

    1984-01-01

    It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

  14. [Experimental research of gaits based on young plantar pressure test].

    Science.gov (United States)

    Meng, Qingyun; Tan, Shili; Yu, Hongliu; Shen, Lixing; Zhuang, Jianhai; Wang, Jinwu

    2014-10-01

    The present paper is to study the center line of the plantar pressure of normal young people, and to find the relation between center line of the plantar pressure and gait stability and balance. The paper gives the testing principle and calculating methods for geometric center of plantar pressure distribution and the center of pressure due to the techniques of footprint frame. The calculating formulas in both x direction and y direction are also deduced in the paper. In the experiments carried out in our laboratory, the gait parameters of 131 young subjects walking as usual speed were acquired, and 14 young subjects of the total were specially analyzed. We then provided reference data for the walking gait database of young people, including time parameters, space parameters and plantar pressure parameters. We also obtained the line of geometry center and pressure center under the foot. We found that the differences existed in normal people's geometric center line and the pressure center line. The center of pressure trajectory revealed foot movement stability. The length and lateral changes of the center line of the plantar pressure could be applied to analysis of the plantar pressure of all kinds of people. The results in this paper are useful in clinical foot disease diagnosis and evaluation of surgical effect.

  15. Safety Evaluation Report: Development of Improved Composite Pressure Vessels for Hydrogen Storage, Lincoln Composites, Lincoln, NE, May 25, 2010

    Energy Technology Data Exchange (ETDEWEB)

    Fort, III, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kallman, Richard A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maes, Miguel [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Skolnik, Edward G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Weiner, Steven C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2010-12-22

    Lincoln Composites operates a facility for designing, testing, and manufacturing composite pressure vessels. Lincoln Composites also has a U.S. Department of Energy (DOE)-funded project to develop composite tanks for high-pressure hydrogen storage. The initial stage of this project involves testing the permeation of high-pressure hydrogen through polymer liners. The company recently moved and is constructing a dedicated research/testing laboratory at their new location. In the meantime, permeation tests are being performed in a corner of a large manufacturing facility. The safety review team visited the Lincoln Composites site on May 25, 2010. The project team presented an overview of the company and project and took the safety review team on a tour of the facility. The safety review team saw the entire process of winding a carbon fiber/resin tank on a liner, installing the boss and valves, and curing and painting the tank. The review team also saw the new laboratory that is being built for the DOE project and the temporary arrangement for the hydrogen permeation tests.

  16. Safety review on unit testing of safety system software of nuclear power plant

    International Nuclear Information System (INIS)

    Liu Le; Zhang Qi

    2013-01-01

    Software unit testing has an important place in the testing of safety system software of nuclear power plants, and in the wider scope of the verification and validation. It is a comprehensive, systematic process, and its documentation shall meet the related requirements. When reviewing software unit testing, attention should be paid to the coverage of software safety requirements, the coverage of software internal structure, and the independence of the work. (authors)

  17. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  18. Test and evaluation of pressure vessel materials

    International Nuclear Information System (INIS)

    Choi, Sun Pil; Hong, Jun Hwa; Nho, Kye Hoe; Han, Dae June; Chi, Se Hwan

    1985-01-01

    We have prepared a method for analyzing the Charpy impact test data, which is deduced from ''the standard anelastic solid equation''. The theoretical expression for the absorbed energy is in a form of W=Wsub(U)+(Wsub(R)-Wsub(U))/ [1+(ωtau) 2 ] showing the Debye characteristics and where tau is given by the Arrhenius equation; tau=tau 0 exp(ΔH/ksub(B)T). Four measurable parameters, at the present stage, can characterize the dynamic hehavior of cracking (Charpy impact result). They are the upper shelf energy(Wsub(R), the lower shelf energy (Wsub(U)), the activation energy of crack (ΔH, and wtau(0) where w tau(0) are the resonance frequency of the specimen and the jumping pre-exponential factor of propagating crack respectively. However the states of R (relaxed) and U (un-relaxed) should be defined from reasonable physical conditions in the future and it is possible that Wsub(U) is small enough to be taken as zero. The effects of irradiation, alloying elements, and heat treatment on the impact results should be interpreted as changes in the above characteristic parameters. The present method has been applied for weld metal of SA 508-2 irradiated up to a fluence of 4x10 18 n/cm 2 , E>1.0Mev, resulting in about 29% decrease in Wsub(R), negligible change in Wsub(U), 5.6 times increase in ωtau 0 , and no change in ΔH. This seems to indicate that irradiation degrades an average value of YOUNG's modulus so that cracks propagate more easily and it does not effect on breaking the lattice bond. However much more systematic analyses should be necessary for correct judgment. It is concluded that the present method is quite adequate for analyzing the Charpy impact data even though plastic deformation in the specimen was not considered separately so that the method should be applied for various cases in order to evaluate the proper trend of effects of irradiation, alloying elements, and heat treatment on the Charpy impact results. (Author)

  19. ATHLET calculations of the pressurizer surge line break (PH-SLB test) at the PMK-2 test facility

    International Nuclear Information System (INIS)

    Krepper, E.; Schaefer, F.

    2000-01-01

    At the Hungarian integral test facility PMK-2 a pressurizer surge line break experiment (PH-SLB test) was carried out with the PHARE 4.2.6b project. The primary objective of the test was to provide experimental data for a surge line break transient at VVER-440 reactors with reduced injection from the emergency core cooling systems (ECC). At the Institute of Safety Research calculations of the experiment were performed with the thermohydraulic computer code ATHLET, which was developed by GRS (Gesellschaft fuer Anlagen- und Reaktorsicherheit) mbH. In the context of the PHARE 4.2.6b project the Institute of Safety Research has also supplied the void fraction measurement system for the PMK-2 test facility and was involved in the evaluation of the experimental results. (orig.)

  20. Technique for unit testing of safety software verification and validation

    International Nuclear Information System (INIS)

    Li Duo; Zhang Liangju; Feng Junting

    2008-01-01

    The key issue arising from digitalization of the reactor protection system for nuclear power plant is how to carry out verification and validation (V and V), to demonstrate and confirm the software that performs reactor safety functions is safe and reliable. One of the most important processes for software V and V is unit testing, which verifies and validates the software coding based on concept design for consistency, correctness and completeness during software development. The paper shows a preliminary study on the technique for unit testing of safety software V and V, focusing on such aspects as how to confirm test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed here was successfully used in the work of unit testing on safety software of a digital reactor protection system. (authors)

  1. Forklift Safety Fundamentals, #20299, Test 20300

    Energy Technology Data Exchange (ETDEWEB)

    Grogin, Phillip W. [Los Alamos National Laboratory

    2017-02-09

    A powered industrial truck (PIT) is defined as a mobile, powerdriven vehicle used to carry, push, pull, lift, or stack material (not including vehicles intended primarily for earth moving). There are many types of and names for PITs, including forklifts, trucks, fork trucks, platform lift trucks, motorized hand trucks, and tractors. Although not every PIT is a forklift, because PITs are commonly called “forklifts,” this course manual generally uses the term “forklift,” although at times the terms “truck” and “PIT” are also used. In some areas of this course, you will see green boxes that refer to the Occupational Safety and Health Administration (OSHA) regulation for PITs, which is 29 Code of Federal Regulations (CFR) 1910.178, Powered Industrial Trucks. The letter in the parentheses refers to the specific section of the regulation.

  2. Safety prediction for basic components of safety critical software based on static testing

    International Nuclear Information System (INIS)

    Son, H.S.; Seong, P.H.

    2001-01-01

    The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, both of which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

  3. Safety prediction for basic components of safety-critical software based on static testing

    International Nuclear Information System (INIS)

    Son, H.S.; Seong, P.H.

    2000-01-01

    The purpose of this work is to develop a safety prediction method, with which we can predict the risk of software components based on static testing results at the early development stage. The predictive model combines the major factor with the quality factor for the components, which are calculated based on the measures proposed in this work. The application to a safety-critical software system demonstrates the feasibility of the safety prediction method. (authors)

  4. Advances in the analysis of pressure interference tests

    Energy Technology Data Exchange (ETDEWEB)

    Martinez R, N. [Petroleos Mexicanos, PEMEX, Mexico City (Mexico); Samaniego V, F. [Univ. Nacional Autonoma de Mexico (Mexico)

    2010-12-15

    This paper presented an extension for radial, linear, and spherical flow conditions of the El-Khatib method for analyzing pressure interference tests through utilization of the pressure derivative. Conventional analysis of interference tests considers only radial flow, but some reservoirs have physical field conditions in which linear or spherical flow conditions prevail. The INTERFERAN system, a friendly computer code for the automatic analysis of pressure interference tests, was also discussed and demonstrated by way of 2 field cases. INTERFERAN relies on the principle of superposition in time and space to interpret a test of several wells with variable histories of production or injection or both. The first field case addressed interference tests conducted in the naturally fractured geothermal field of Klamath Falls, and the second field case was conducted in a river-formed bed in which linear flow conditions are dominant. The analysis was deemed to be reliable. 13 refs., 1 tab., 7 figs.

  5. Minnesota urban partnership agreement national evaluation : safety data test plan.

    Science.gov (United States)

    2009-11-17

    This report provides the safety data test plan for the Minnesota Urban Partnership Agreement (UPA) under the United States Department of Transportation (U.S. DOT) UPA Program. The Minnesota UPA projects focus on reducing congestion by employing strat...

  6. Safety of nuclear pressure vessels and its regulatory aspects in France

    Energy Technology Data Exchange (ETDEWEB)

    de Torquat, G; Queniart, D; Barrachin, B; Roche, R

    1979-01-01

    Having outlined the basic French regulations governing the safety of both pressure vessels and also of nuclear installations in general the particular safety regulations covering prestressed concrete vessels for nuclear reactors are considered. The regulations now being prepared to cover heat transfer systems of water reactors are detailed under sections headed; general provisions, sizing, and construction.

  7. Test interval optimization of safety systems of nuclear power plant using fuzzy-genetic approach

    International Nuclear Information System (INIS)

    Durga Rao, K.; Gopika, V.; Kushwaha, H.S.; Verma, A.K.; Srividya, A.

    2007-01-01

    Probabilistic safety assessment (PSA) is the most effective and efficient tool for safety and risk management in nuclear power plants (NPP). PSA studies not only evaluate risk/safety of systems but also their results are very useful in safe, economical and effective design and operation of NPPs. The latter application is popularly known as 'Risk-Informed Decision Making'. Evaluation of technical specifications is one such important application of Risk-Informed decision making. Deciding test interval (TI), one of the important technical specifications, with the given resources and risk effectiveness is an optimization problem. Uncertainty is inherently present in the availability parameters such as failure rate and repair time due to the limitation in assessing these parameters precisely. This paper presents a solution to test interval optimization problem with uncertain parameters in the model with fuzzy-genetic approach along with a case of application from a safety system of Indian pressurized heavy water reactor (PHWR)

  8. Three-Dimensional Digital Image Correlation of a Composite Overwrapped Pressure Vessel During Hydrostatic Pressure Tests

    Science.gov (United States)

    Revilock, Duane M., Jr.; Thesken, John C.; Schmidt, Timothy E.

    2007-01-01

    Ambient temperature hydrostatic pressurization tests were conducted on a composite overwrapped pressure vessel (COPV) to understand the fiber stresses in COPV components. Two three-dimensional digital image correlation systems with high speed cameras were used in the evaluation to provide full field displacement and strain data for each pressurization test. A few of the key findings will be discussed including how the principal strains provided better insight into system behavior than traditional gauges, a high localized strain that was measured where gages were not present and the challenges of measuring curved surfaces with the use of a 1.25 in. thick layered polycarbonate panel that protected the cameras.

  9. A single-stage high pressure steam injector for next generation reactors: test results and analysis

    International Nuclear Information System (INIS)

    Cattadori, G.; Galbiati, L.; Mazzocchi, L.; Vanini, P.

    1995-01-01

    Steam injectors can be used in advanced light water reactors (ALWRs) for high pressure makeup water supply; this solution seems to be very attractive because of the ''passive'' features of steam injectors, that would take advantage of the available energy from primary steam without the introduction of any rotating machinery. The reference application considered in this work is a high pressure safety injection system for a BWR; a water flow rate of about 60 kg/s to be delivered against primary pressures covering a quite wide range up to 9 MPa is required. Nevertheless, steam driven water injectors with similar characteristics could be used to satisfy the high pressure core coolant makeup requirements of next generation PWRs. With regard to BWR application, an instrumented steam injector prototype with a flow rate scaling factor of about 1:6 has been built and tested. The tested steam injector operates at a constant inlet water pressure (about 0.2 MPa) and inlet water temperature ranging from 15 to 37 o C, with steam pressure ranging from 2.5 to 8.7 MPa, always fulfilling the discharge pressure target (10% higher than steam pressure). To achieve these results an original double-overflow flow rate-control/startup system has been developed. (Author)

  10. Pressure Distribution Tests on a Series of Clark Y Biplane Cellules with Special Reference to Stability

    Science.gov (United States)

    Noyes, Richard W

    1933-01-01

    The pressure distribution data discussed in this report represents the results of part of an investigation conducted on the factors affecting the aerodynamic safety of airplanes. The present tests were made on semispan, circular-tipped Clark Y airfoil models mounted in the conventional manner on a separation plane. Pressure readings were made simultaneously at all test orifices at each of 20 angles of attack between -8 degrees and +90 degrees. The results of the tests on each wing arrangement are compared on the bases of maximum normal force coefficient, lateral stability at a low rate of roll, and relative longitudinal stability. Tabular data are also presented giving the center of pressure location of each wing.

  11. Safety grade pressurizer heater power supply connector assembly

    International Nuclear Information System (INIS)

    Burnett, J.M.; Daftari, R.M.; Reyns, R.M.

    1987-01-01

    This patent describes a pressurizer heater power supply connector assembly for attaching a power cable to an electric heater within a pressurizer of a pressurized water nuclear reactor system, the electric heater having pin contacts. The assembly comprises: a pin-socket type connector including a tubular body having a first open end carrying a pin-socket contact member and an insert intermediate a shell and the pin-socket contact member, the contact member having socket means for electrically receiving and contacting the pin contacts, and a second open end; a flexible sealed conduit including a flexible corrugated tube having one end connected to the second open end of the pin-socket type connector, and another end; and a shop splice assembly including a header adapter and a hose clamp interconnected between the header adapter and another end of the flexible corrugated tube

  12. Automated corrosion fatigue crack growth testing in pressurized water environments

    International Nuclear Information System (INIS)

    Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

    1984-01-01

    This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

  13. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  14. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  15. Contribution of NDT to the safety of pressurized components in power stations

    International Nuclear Information System (INIS)

    Mletzko, U.; Maier, H.J.

    1994-01-01

    In the eyes on the MPA Stuttgart, the nondestructive testing has a very high weight relating to the safety of pressure components in power stations (concept of basis safety). In this connection, the performance verification by NDT has a special significance. A qualification of NDT-techniques can be, indeed, executed in the initial stage at test bodies with artificial faults, known with respect to position, size and type. Even theoretical (modelling) considerations can be integrated. For a performance verification in a closer sense however, this is not sufficient. The performance verification should be effected for the overall system, composed of hardware, software and examination personnel at components having the scale of 1:1 (Full Scale) under realistic boundary conditions and given times. The components must have natural or quasi-natural faults in a certain quality. The informative value of performance verifications is considerably increased, when executed as authenitic dry runs, and when the fault state is subsequently verified by destructive (metallographical) methods. (orig.) [de

  16. Well control during the drilling and testing of high pressure offshore wells

    Energy Technology Data Exchange (ETDEWEB)

    1992-05-01

    This Code has been prepared for use as a guide to safe practice for those concerned with well control during the drilling and testing of high pressure offshore wells. It is intended to provide information and guidance on those well control activities associated with high pressure wells which have an impact on safety offshore, and therefore require detailed care and attention. The Code has been produced in a United Kingdom Continental Shelf (UKCS) context, but the principles and recommendations have general relevance to similar operations elsewhere. (author)

  17. High-pressure test loop design and application

    International Nuclear Information System (INIS)

    Burnette, R.D.; Graves, J.N.; Blair, P.G.; Baldwin, N.L.

    1980-07-01

    A high-pressure test loop (HPTL) has been constructed for the purpose of performing a number of chemistry experiments at simulated HTGR conditions of temperature, pressure, flow, and impurity content. The HPTL can be used to develop, modify, and verify computer codes for a variety of chemical processes involving gas phase transport in the reactor. Processes such as graphite oxidation, fission product transport, fuel reactions, purification systems, and dust entrainment can be studied at high pressure, which would largely eliminate difficulties in correlating existing laboratory data and reactor conditions

  18. Safety test facilities - status, needs, future directions

    International Nuclear Information System (INIS)

    Heusener, G.; Cogne, F.

    1979-08-01

    A survey is given of the in-pile programs which are presently or in the near future being performed in the DeBeNe-area and in France. Only those in-pile programs are considered which are dealing with severe accidents that might lead to disruption of major parts of the core. By comparing the needs with the goals of the present programs points are identified which are not sufficiently well covered up till now. The future procedure is described: the existing facilities will be used to the largest possible extent. Whenever it is necessary, upgrading and improvement will be foreseen. Studies of a Test Facility allowing the transient testing of large pin bundles should be continued. The construction of such a facility in Europe in the near future however seems premature

  19. Safety assessment for the rf Test Facility

    International Nuclear Information System (INIS)

    Nagy, A.; Beane, F.

    1984-08-01

    The Radio Frequency Test Facility (RFTF) is a part of the Magnetic Fusion Program's rf Heating Experiments. The goal of the Magnetic Fusion Program (MFP) is to develop and demonstrate the practical application of fusion. RFTF is an experimental device which will provide an essential link in the research effort aiming at the realization of fusion power. This report was compiled as a summary of the analysis done to ensure the safe operation of RFTF

  20. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzed in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR safety

  1. Common safety approach for future pressurized water reactors in France and Germany

    International Nuclear Information System (INIS)

    Queniart, D.; Gros, G.

    1994-01-01

    In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. This holds for utilities and industry in the development of a joint PWR project, the ''European Pressurized Water Reactor (EPR)'' by Electricite de France (EDF), German utilities, Nuclear Power International (NPI), Framatome and Siemens as well as for the technical safety objectives for future evolutionary reactors on the basis of a common safety approach adopted by the safety authorities of both countries for plants to operate form the beginning of the next century. The proposed paper covers this common development of a safety approach and particular technical safety objectives. (authors). 5 refs. 1 fig

  2. Imaging indicator for ESD safety testing.

    Energy Technology Data Exchange (ETDEWEB)

    Whinnery, LeRoy L.,; Nissen, April; Keifer, Patrick N.; Tyson, Alexander

    2013-05-01

    This report describes the development of a new detection method for electrostatic discharge (ESD) testing of explosives, using a single-lens reflex (SLR) digital camera and a 200-mm macro lens. This method has demonstrated several distinct advantages to other current ESD detection methods, including the creation of a permanent record, an enlarged image for real-time viewing as well as extended periods of review, and ability to combine with most other Go/No-Go sensors. This report includes details of the method, including camera settings and position, and results with wellcharacterized explosives PETN and RDX, and two ESD-sensitive aluminum powders.

  3. Pressure Drop Test of Hybrid Mixing Vane Spacer Grid

    Energy Technology Data Exchange (ETDEWEB)

    Oh, D. S.; Chang, S. K.; Kim, B. D.; Chun, S. Y.; Chun, T. H

    2007-08-15

    The pressure loss test has been accomplished in the test section containing 5x5 rod bundle with a length of 2 m including 3 spacer grids. The test has been performed for the 5 kinds of spacer grids to compare the pressure loss characteristics: 1. Plain spacer grid which has the same body of the Hybrid but without vane (Plain), 2. Hybrid Vane spacer grid (Hybrid), 3. Hybrid-SC spacer grid which is constructed with coined, chamfered strip and is fabricated by spot welding, 4. Hybrid-LC spacer grid which is constructed with coined, chamfered strip and is fabricated by line welding along intersection line, 5. Westinghouse spacer grid with split vane (Plus-7). The pressure loss coefficient of the Plain, Hybrid, Hybrid-SC, Hybrid-LC, and Plus-7 spacer grid is 0.93, 1.15, 1.02, 1.04, and 1.08, respectively.

  4. Well control during the drilling and testing of high pressure offshore wells

    Energy Technology Data Exchange (ETDEWEB)

    1992-05-01

    This Code has been prepared for use as a guide to safe practice for those concerned with well control during the drilling and testing of high pressure offshore wells. It is intended to provide information and guidance on those well control activities associated with high pressure wells which have an impact on safety offshore, and therefore require detailed care and attention. The Code has been produced in a United Kingdom Continental Shelf (UKCS) context, but the principles and recommendations have general relevance to similar operations elsewhere. Each chapter of the Code covers an important aspect of well control and has an introduction which describes the part each activity plays in the drilling and testing of high pressure offshore wells. (Author)

  5. Gas reactor in-pile safety test project (GRIST-2)

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Arbtin, E.; St Pierre, R.

    1979-01-01

    Although out-of-pile tests may be expected to confirm individual phenomena models in core disruptive accident analysis codes, only in-pile tests are capable of verifying the extremely complex integrated model effects within the appropriate time phase for these accidents. For this reason, the GRIST-2 project, the purpose of which is to design and construct an in-pile helium loop capable of transient safety testing in the TREAT facility in Idaho, forms a cornerstone of the US GCFR safety program. The project organization, experiment program, facility, helium system design, and schedule which have been selected to meet the objectives are described

  6. A Study on the Improvement of Safety Testing Standards and Methods for Mammography

    International Nuclear Information System (INIS)

    Choi, Seon Hyeong; Jung, Ah Young; Yong, Hwan Seok; Kim, Do Wan; Jang, Gi Won; Cha, Sang Hoon; Jo, Sang Won; Park, Ji Koon

    2012-01-01

    To establish the improved national safety testing standards and methods for mammography. We investigated and compared the current status of mammographic equipment installation with the national and international safety and quality control programs and methods. We established and verified the draft for safety testing standards and methods. We propose that the investigations of the conductor system, hardware leakage radiation profile, illumination intensity test, comparison between X-ray and light photon exposure, X-ray dose exposure on the chest wall, compression equipment size, timing equipment, and the average effective radiation dose, should all be maintained as they are in the present state without any changes. However, the exposure radiation dose reproducibility, kVp and mAs, and the half value layer tests should be reconsidered and revised. Moreover, compression pressure and autonomic exposure control system (AEC) tests should be included as new criteria. Other parameter controls included in the phantom image analysis which overlap with total quality assurance should be excluded. We recommend that AEC and compression pressure tests should be included as new criteria and the methods for the exposure radiation dose reproducibility, kVp, and mAs, and half value layer tests should be reconsidered and revised.

  7. G-Tunnel pressurized slot-testing preparations

    International Nuclear Information System (INIS)

    Zimmerman, R.M.; Sifre-Soto, C.; Mann, K.L.; Bellman, R.A. Jr.; Luker, S.; Dodds, D.J.

    1992-04-01

    Designers and analysts of radioactive waste repositories must be able to predict the mechanical behavior of the host rock. Sandia National laboratories elected to conduct a development program on pressurized slot testing and featured (1) development of an improved method to cut slots using a chain saw with diamond-tipped cutters, (2) measurements useful for determining in situ stresses normal to slots, (3) measurements applicable for determining the in situ modulus of deformation parallel to a drift surface, and (4) evaluations of the potentials of pressurized slot strength testing. This report describes the preparations leading to the measurements and evaluations

  8. The oil pressure test of the hydraulic impeller blade

    Science.gov (United States)

    Ye, Wen-bo; Jia, Li-tao

    2017-12-01

    This article introduced the structure of the Kaplan runner in hydropower station and the operating process of the oil pressure test has been described. What’s more, the whole process, including filling oil to the runner hub, the movement of the runner blade, the oil circuit, have been presented in detail.Since the manipulation of the oil circuit which controlled by three Valve groups consisting of six valves was complicated, the author is planning to replace them with 3-position 3-way electromagnetic valves, so we can simplify the operation procedure.The author hopes this article can provide technical reference for the oil pressure test.

  9. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  10. Safety tests carried out at Cadarache. Sodium fires

    International Nuclear Information System (INIS)

    Fruchard, M.

    1976-01-01

    Safety test on sodium fires developed at the Cadarache Nuclear Centre by the Department of Nuclear Safety, section of safety experiments on radioactivity transfer are conducted in two main directions: analysis of the behavior and thermodynamic consequences of accidental fires, working on the basis of typical experimental results; research and development of methods and equipment to control and if possible extinguish these fires. The most important part of this programme is concerned with the sodium pool fires which would result from the failure of a secondary coolant circuit pipe [fr

  11. Toward an understanding of the impact of production pressure on safety performance in construction operations.

    Science.gov (United States)

    Han, Sanguk; Saba, Farzaneh; Lee, Sanghyun; Mohamed, Yasser; Peña-Mora, Feniosky

    2014-07-01

    It is not unusual to observe that actual schedule and quality performances are different from planned performances (e.g., schedule delay and rework) during a construction project. Such differences often result in production pressure (e.g., being pressed to work faster). Previous studies demonstrated that such production pressure negatively affects safety performance. However, the process by which production pressure influences safety performance, and to what extent, has not been fully investigated. As a result, the impact of production pressure has not been incorporated much into safety management in practice. In an effort to address this issue, this paper examines how production pressure relates to safety performance over time by identifying their feedback processes. A conceptual causal loop diagram is created to identify the relationship between schedule and quality performances (e.g., schedule delays and rework) and the components related to a safety program (e.g., workers' perceptions of safety, safety training, safety supervision, and crew size). A case study is then experimentally undertaken to investigate this relationship with accident occurrence with the use of data collected from a construction site; the case study is used to build a System Dynamics (SD) model. The SD model, then, is validated through inequality statistics analysis. Sensitivity analysis and statistical screening techniques further permit an evaluation of the impact of the managerial components on accident occurrence. The results of the case study indicate that schedule delays and rework are the critical factors affecting accident occurrence for the monitored project. Copyright © 2013 Elsevier Ltd. All rights reserved.

  12. Safety analyses for an in-pile SCWR fuel qualification test loop

    Energy Technology Data Exchange (ETDEWEB)

    Schulenberg, T.; Raque, M. [Karlsruhe Inst. of Tech., Karlsruhe (Germany)

    2014-07-01

    As a nuclear facility cooled with supercritical water has never been built nor operated in the past, the planned SCWR fuel qualification test will give the first experience with supercritical water-cooled nuclear systems in general. With a fuel inventory of almost 1 kg of UO{sub 2} with almost 20% enrichment, the supercritical pressure test section inside a low pressure, pool type research reactor needs to be cooled properly even in case of a number of postulated design basis accidents. Depressurization systems and emergency cooling systems will need to be designed with similar reliability as for a prototype reactor to ensure the integrity of barriers retaining the radioactive material. The paper reports about the safety concept and summarizes the safety analyses which have been performed in this context. (author)

  13. Temperature effect compensation for fast differential pressure decay testing

    International Nuclear Information System (INIS)

    Shi, Yan; Tong, Xiaomeng; Cai, Maolin

    2014-01-01

    To avoid the long temperature recovery period with differential pressure decay for leak detection, a novel method with temperature effect compensation is proposed to improve the testing efficiency without full stabilization of temperature. The mathematical model of conventional differential pressure decay testing is established to analyze the changes of temperature and pressure during the measuring period. Then the differential pressure is divided into two parts: the exponential part caused by temperature recovery and the linear part caused by leak. With prior information obtained from samples, parameters of the exponential part can be identified precisely, and the temperature effect will be compensated before it fully recovers. To verify the effect of the temperature compensated method, chambers with different volumes are tested under various pressures and the experiments show that the improved method is faster with satisfactory precision, and an accuracy less than 0.25 cc min −1  can be achieved when the compensation time is proportional to four times the theoretical thermal-time constant. (paper)

  14. Test Results of a Platform for Safety I and C Systems of SMART MMIS

    International Nuclear Information System (INIS)

    Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong

    2011-01-01

    SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, has been under development at KAERI since 1997. A standard design safety analysis report of the SMART prepared by KAERI was submitted to Korea institute of nuclear safety (KINS) at the end of 2010. KAERI aims to achieve standard design approval (SDA) from KINS by the end of 2011. SMART MMIS has been designed using digital systems. It has digital-based compact control rooms. Its instrumentation and control (I and C) systems are designed using modular equipment connected through datalinks. Non-safety I and C systems are designed based on the commercial distributed control systems. Safety I and C systems are based on a new platform developed by KAERI. The platform is a high-speed digital signal processor (DSP)-based control unit. It plays the role of a module that provides control functions of the safety I and C systems. The test facilities have been developed at KAERI since 2009. This paper presents the development and test results of the platform

  15. The safety related aspects of pressure components in nuclear power plants

    International Nuclear Information System (INIS)

    Lindackers, K.H.

    1979-01-01

    Over the last two years the safety philosophy for nuclear power plants in the Federal Republic of Germany has changed considerably, as everyone working in the field perceives. The original and appropriate philosophy of risk minimalisation through graduated safety barriers has been more and more replaced by the utopian goal of total prevention of any damage. The reasons for this development are discussed briefly especially regarding pressure components. The very numerous pressure components of a nuclear power station are not all of equal importance with respect to safety. Although considerable efforts have been made, it has not been possible, to date, to achieve an agreement between operators, manufacturers, licensing authorities, independent experts, and other specialists about the safety related classification of the manifold pressure bearing parts in nuclear power stations. The background of this extremely regrettable situation is explained. In the last part of the paper the author suggests a simple and clear safety philosophy for pressure components in nuclear power stations. This philosophy is orientated both on Safety Regulations of the Radiation Protection Decree ('Strahlenschutzverordnung') of the 13th October 1976 and on the Safety Criteria for Nuclear Power Stations from 21st October 1977. Only a simple, clear framework can make a contribution to the further improvement of the already exceptional safety of nuclear facilities and to the removal of obstacles in the licensing procedure which, taken as a whole, tie up skilled personnel to a senseless degree, involve considerable financial expenditure, and have no relevance for the safety of nuclear power plants. (orig.) [de

  16. Delayed hydrogen cracking test design for pressure tubes

    International Nuclear Information System (INIS)

    Haddad, Roberto; Loberse, Antonio N.; Yawny, Alejandro A.; Riquelme, Pablo

    1999-01-01

    CANDU nuclear power stations pressure tubes of alloy Zr-2,5 % Nb present a cracking phenomenon known as delayed hydrogen cracking (DHC). This is a brittle fracture of zirconium hydrides that are developed by hydrogen due to aqueous corrosion on the metal surface. This hydrogen diffuses to the crack tip where brittle zirconium hydrides develops and promotes the crack propagation. A direct current potential decay (DCPD) technique has been developed to measure crack propagation rates on compact test (CT) samples machined from a non irradiated pressure tube. Those test samples were hydrogen charged by cathodic polarization in an acid solution and then pre cracked in a fatigue machine. This technique proved to be useful to measure crack propagation rates with at least 1% accuracy for DHC in pressure tubes. (author)

  17. Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

    International Nuclear Information System (INIS)

    Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

    1979-01-01

    The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

  18. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Ancuta, M.; Radu, V.; Stefan, V.; Preda, M.

    2001-01-01

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 10 21 n·cm -2 ; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (E n > 1 MeV), 1 - 2 · 10 13 ncm -2 s -1 ; - neutron fluence (E n > 1 MeV), 4 · 10 20 ncm -2 . The following characteristics were obtained from tensile

  19. 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)

  20. Pressure-Sensitive Paints Advance Rotorcraft Design Testing

    Science.gov (United States)

    2013-01-01

    The rotors of certain helicopters can spin at speeds as high as 500 revolutions per minute. As the blades slice through the air, they flex, moving into the wind and back out, experiencing pressure changes on the order of thousands of times a second and even higher. All of this makes acquiring a true understanding of rotorcraft aerodynamics a difficult task. A traditional means of acquiring aerodynamic data is to conduct wind tunnel tests using a vehicle model outfitted with pressure taps and other sensors. These sensors add significant costs to wind tunnel testing while only providing measurements at discrete locations on the model's surface. In addition, standard sensor solutions do not work for pulling data from a rotor in motion. "Typical static pressure instrumentation can't handle that," explains Neal Watkins, electronics engineer in Langley Research Center s Advanced Sensing and Optical Measurement Branch. "There are dynamic pressure taps, but your costs go up by a factor of five to ten if you use those. In addition, recovery of the pressure tap readings is accomplished through slip rings, which allow only a limited amount of sensors and can require significant maintenance throughout a typical rotor test." One alternative to sensor-based wind tunnel testing is pressure sensitive paint (PSP). A coating of a specialized paint containing luminescent material is applied to the model. When exposed to an LED or laser light source, the material glows. The glowing material tends to be reactive to oxygen, explains Watkins, which causes the glow to diminish. The more oxygen that is present (or the more air present, since oxygen exists in a fixed proportion in air), the less the painted surface glows. Imaged with a camera, the areas experiencing greater air pressure show up darker than areas of less pressure. "The paint allows for a global pressure map as opposed to specific points," says Watkins. With PSP, each pixel recorded by the camera becomes an optical pressure

  1. New method of safety assessment for pressure vessel of nuclear power plant--brief introduction of master curve approach

    International Nuclear Information System (INIS)

    Yang Wendou

    2011-01-01

    The new Master Curve Method is called as a revolutionary advance to the assessment of- reactor pressure vessel integrity in USA. This paper explains the origin, basis and standard of the Master Curve from the reactor pressure-temperature limit curve which assures the safety of nuclear power plant. According to the characteristics of brittle fracture which is greatly susceptible to the microstructure, the theory and the test method of the Master Curve as well as its statistical law which can be modeled using Weibull distribution are described in this paper. The meaning, advantage, application and importance of the Master Curve as well as the relation between the Master Curve and nuclear power safety are understood from the fitting formula for the fracture toughness database by Weibull distribution model. (author)

  2. Experimental validation of a Lyapunov-based controller for the plasma safety factor and plasma pressure in the TCV tokamak

    Science.gov (United States)

    Mavkov, B.; Witrant, E.; Prieur, C.; Maljaars, E.; Felici, F.; Sauter, O.; the TCV-Team

    2018-05-01

    In this paper, model-based closed-loop algorithms are derived for distributed control of the inverse of the safety factor profile and the plasma pressure parameter β of the TCV tokamak. The simultaneous control of the two plasma quantities is performed by combining two different control methods. The control design of the plasma safety factor is based on an infinite-dimensional setting using Lyapunov analysis for partial differential equations, while the control of the plasma pressure parameter is designed using control techniques for single-input and single-output systems. The performance and robustness of the proposed controller is analyzed in simulations using the fast plasma transport simulator RAPTOR. The control is then implemented and tested in experiments in TCV L-mode discharges using the RAPTOR model predicted estimates for the q-profile. The distributed control in TCV is performed using one co-current and one counter-current electron cyclotron heating actuation.

  3. A Test Suite for Safety-Critical Java using JML

    DEFF Research Database (Denmark)

    Ravn, Anders Peter; Søndergaard, Hans

    2013-01-01

    Development techniques are presented for a test suite for the draft specification of the Java profile for Safety-Critical Systems. Distinguishing features are: specification of conformance constraints in the Java Modeling Language, encoding of infrastructure concepts without implementation bias......, and corresponding specifications of implicitly stated behavioral and real-time properties. The test programs are auto-generated from the specification, while concrete values for test parameters are selected manually. The suite is open source and publicly accessible....

  4. Acoustic emission signal measurements in pressure vessel testing

    International Nuclear Information System (INIS)

    Peter, A.

    1984-01-01

    The number of acoustic emission events per plastically deformed unit of volume caused by artificial notches in real pressure vessels has been calculated taking into account reference voltage, distance between acoustic emission source and sensor as well as the effect of noise background. A test performed at a 100 m 3 gasholder verifies the theoretical considerations. (author)

  5. 49 CFR 178.345-13 - Pressure and leakage tests.

    Science.gov (United States)

    2010-10-01

    ... may not prevent the detection of leaks, or damage the device. Restraining devices must be removed....345-13 Transportation Other Regulations Relating to Transportation PIPELINE AND HAZARDOUS MATERIALS... leak tested at not less than 80 percent of tank's MAWP with the pressure maintained for at least 5...

  6. Quantification of Safety-Critical Software Test Uncertainty

    International Nuclear Information System (INIS)

    Khalaquzzaman, M.; Cho, Jaehyun; Lee, Seung Jun; Jung, Wondea

    2015-01-01

    The method, conservatively assumes that the failure probability of a software for the untested inputs is 1, and the failure probability turns in 0 for successful testing of all test cases. However, in reality the chance of failure exists due to the test uncertainty. Some studies have been carried out to identify the test attributes that affect the test quality. Cao discussed the testing effort, testing coverage, and testing environment. Management of the test uncertainties was discussed in. In this study, the test uncertainty has been considered to estimate the software failure probability because the software testing process is considered to be inherently uncertain. A reliability estimation of software is very important for a probabilistic safety analysis of a digital safety critical system of NPPs. This study focused on the estimation of the probability of a software failure that considers the uncertainty in software testing. In our study, BBN has been employed as an example model for software test uncertainty quantification. Although it can be argued that the direct expert elicitation of test uncertainty is much simpler than BBN estimation, however the BBN approach provides more insights and a basis for uncertainty estimation

  7. Design and testing of high-pressure railguns and projectiles

    International Nuclear Information System (INIS)

    Peterson, D.R.; Fowler, C.M.

    1984-01-01

    The results of high-pressure tests of four railgun designs and four projectile types are presented. All tests were conducted at the Los Alamos explosive magnetic-flux compression facility in Ancho Canyon. The data suggest that the high-strength projectiles have lower resistance to acceleration than the low strength projectiles, which expand against the bore during acceleration. The railguns were powered by explosive magneticflux compression generators. Calculations to predict railgun and power supply performance were performed by Kerrisk

  8. Acceptance test report, plutonium finishing plant life safety upgrade

    International Nuclear Information System (INIS)

    Hodge, S.G.

    1994-01-01

    This acceptance Test Procedure (ATP) has been prepared to demonstrate that modifications to the Fir Protection systems function as required by project criteria. The ATP will test the Fire Alarm Control Panels, Flow Alarm Pressure Switch, Heat Detectors, Smoke Detectors, Flow Switches, Manual Pull Stations, and Gong/Door By Pass Switches

  9. Status of the EU test blanket systems safety studies

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-01-01

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  10. Status of the EU test blanket systems safety studies

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-10-15

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  11. Common safety approach for future pressurized water reactors in France and Germany

    International Nuclear Information System (INIS)

    Frisch, W.; Jahns, A.; Queniart, D.; Gros, G.

    1995-01-01

    In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. The proposed paper covers this common development of a safety approach and particular technical safety objectives. The main topics of this document are presented in this paper, together with a rationale for the approach and the recommended technical principles. (K.A.). 5 refs., 1 fig

  12. Safety Test Program Summary SNAP 19 Pioneer Heat Source Safety Program

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1971-07-01

    Sixteen heat source assemblies have been tested in support of the SNAP 19 Pioneer Safety Test Program. Seven were subjected to simulated reentry heating in various plasma arc facilities followed by impact on earth or granite. Six assemblies were tested under abort accident conditions of overpressure, shrapnel impact, and solid and liquid propellant fires. Three capsules were hot impacted under Transit capsule impact conditions to verify comparability of test results between the two similar capsule designs, thus utilizing both Pioneer and Transit Safety Test results to support the Safety Analysis Report for Pioneer. The tests have shown the fuel is contained under all nominal accident environments with the exception of minor capsule cracks under severe impact and solid fire environments. No catastrophic capsule failures occurred in this test which would release large quantities of fuel. In no test was fuel visible to the eye following impact or fire. Breached capsules were defined as those which exhibit thoria contamination on its surface following a test, or one which exhibited visible cracks in the post test metallographic analyses.

  13. Module Testing Techniques for Nuclear Safety Critical Software Using LDRA Testing Tool

    International Nuclear Information System (INIS)

    Moon, Kwon-Ki; Kim, Do-Yeon; Chang, Hoon-Seon; Chang, Young-Woo; Yun, Jae-Hee; Park, Jee-Duck; Kim, Jae-Hack

    2006-01-01

    The safety critical software in the I and C systems of nuclear power plants requires high functional integrity and reliability. To achieve those requirement goals, the safety critical software should be verified and tested according to related codes and standards through verification and validation (V and V) activities. The safety critical software testing is performed at various stages during the development of the software, and is generally classified as three major activities: module testing, system integration testing, and system validation testing. Module testing involves the evaluation of module level functions of hardware and software. System integration testing investigates the characteristics of a collection of modules and aims at establishing their correct interactions. System validation testing demonstrates that the complete system satisfies its functional requirements. In order to generate reliable software and reduce high maintenance cost, it is important that software testing is carried out at module level. Module testing for the nuclear safety critical software has rarely been performed by formal and proven testing tools because of its various constraints. LDRA testing tool is a widely used and proven tool set that provides powerful source code testing and analysis facilities for the V and V of general purpose software and safety critical software. Use of the tool set is indispensable where software is required to be reliable and as error-free as possible, and its use brings in substantial time and cost savings, and efficiency

  14. NRC confirmatory safety system testing in support of AP600 design review

    International Nuclear Information System (INIS)

    Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.

    1994-01-01

    Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation

  15. Acceptance test procedure for High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    The overall objective of the acceptance test is to demonstrate a combined system. This includes associated tools and equipment necessary to perform cleaning in the 105 K East Basin (KE) for achieving optimum reduction in the level of contamination/dose rate on canisters prior to removal from the KE Basin and subsequent packaging for disposal. Acceptance tests shall include necessary hardware to achieve acceptance of the cleaning phase of canisters. This acceptance test procedure will define the acceptance testing criteria of the high pressure water jet cleaning fixture. The focus of this procedure will be to provide guidelines and instructions to control, evaluate and document the acceptance testing for cleaning effectiveness and method(s) of removing the contaminated surface layer from the canister presently identified in KE Basin. Additionally, the desired result of the acceptance test will be to deliver to K Basins a thoroughly tested and proven system for underwater decontamination and dose reduction. This report discusses the acceptance test procedure for the High Pressure Water Jet

  16. Safety and Function Test Report for the SWIFT Wind Turbine

    Energy Technology Data Exchange (ETDEWEB)

    Mendoza, I.; Hur, J.

    2013-01-01

    This test was conducted as part of the U.S. Department of Energy's (DOE) Independent Testing project. This project was established to help reduce the barriers of wind energy expansion by providing independent testing results for small turbines. Three turbines where selected for testing at the National Wind Technology Center (NWTC) as a part of round two of the Small Wind Turbine Independent Testing project. Safety and Function testing is one of up to 5 tests that may be performed on the turbines. Other tests include power performance, duration, noise, and power quality. The results of the testing will provide the manufacturers with reports that may be used for small wind turbine certification.

  17. Development test procedure High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    Development testing will be performed on the water jet cleaning fixture to determine the most effective arrangement of water jet nozzles to remove contamination from the surfaces of canisters and other debris. The following debris may be stained with dye to simulate surface contaminates: Mark O, Mark I, and Mark II Fuel Storage Canisters (both stainless steel and aluminum), pipe of various size, (steel, stainless, carbon steel and aluminum). Carbon steel and stainless steel plate, channel, angle, I-beam and other surfaces, specifically based on the Scientific Ecology Group (SEG) inventory and observations of debris within the basin. Test procedure for developmental testing of High Pressure Water Jet System

  18. Evaluation of high pressure Freon decontamination. I. Preliminary tests

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1983-01-01

    High-pressure Freon blasting techniques are being evaluated for applications involving the removal of non-adherent radioactive particulate contamination at SRP. Very little waste is generated by this technique because the used Freon can be easily distilled and reused. One of the principle advantages of this technique is that decontaminated electrical equipment can be returned to service immediately without drying, unlike high-pressure water blasting techniques. Preliminary scoutin tests evaluating high-pressure Freon blasting for decontamination at SRP were carried out at Quadrex Co., Oak Ridge, TN, October 12 and 13. DWPF-type contamination (raw sludge plus volatiles) and separations area-type contamination (diluted boiling point (47.6 0 C) allow it to rapidly separate from higher boiling contaminants via distillation with filtration to remove particulate material, and distillation with condensation, the solvent may be recovered for indefinite reuse while reducing the radioactive waste to a minimum. 3 references, 5 figures, 6 tables

  19. Manipulator for testing a top-opened reactor pressure vessel

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

    The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

  20. Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

    International Nuclear Information System (INIS)

    Atkinson, S.A.

    1989-01-01

    The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab

  1. Safety demonstration test on solvent fire in fuel reprocessing plant

    International Nuclear Information System (INIS)

    Nishio, Gunji; Hashimoto, Kazuichiro

    1989-03-01

    This report summarizes a fundamental of results obtained in the Reprocessing Plant Safety Demonstration Test Program which was performed under the contract between the Science and Technology Agency of Japan and the Japan Atomic Energy Research Institute. In this test program, a solvent fire was hypothesized, and such data were obtained as fire behavior, smoke behavior and integrity of exhaust filters in the ventilation system. Through the test results, it was confirmed that under the fire condition in hypothetical accident, the integrity of the cell and the cell ventilation system were maintained, and the safety function of the exhaust filters was maintained against the smoke loading. Analytical results by EVENT code agreed well with the present test data on the thermofluid flow in a cell ventilation system. (author)

  2. Neutron Irradiation Tests of Pressure Transducers in Liquid Helium

    CERN Document Server

    Amand, J F; Casas-Cubillos, J; Thermeau, J P

    1999-01-01

    The superconducting magnets of the future Large Hadron Collider (LHC) at CERN will operate in pressurised superfluid helium (1 bar, 1.9 K). About 500 pressure transducers will be placed in the liquid helium bath for monitoring the filling and the pressure transients after resistive transitions. Their precision must remain better than 100 mbar at pressures below 2 bar and better than 5% for higher pressures (up to 20 bar), with temperatures ranging from 1.8 K to 300 K. All the tested transducers are based on the same principle: the fluid or gas is separated from a sealed reference vacuum by an elastic membrane; its deformation indicates the pressure. The transducers will be exposed to high neutron fluence (2 kGy, 1014 n/cm2 per year) during the 20 years of machine operation. This irradiation may induce changes both on the membranes characteristics (leakage, modification of elasticity) and on gauges which measure their deformations. To investigate these effects and select the transducer to be used in the LHC, a...

  3. Testing of low pressure proton exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Bettoni, M; Naso, V; Lucentini, M; Rubini, L

    1998-07-01

    One of the main issues concerning PEMFC is the choice of operating pressure, for both stationary and automotive applications. This is because the air compressor may absorb a significant amount--up to 25%--of the power output of the fuel cells stack. A comparison has been made between the performance of various stacks of different dimensions, tested in the De Nora Laboratories operated at high (4 bar) and low (1.5 bar) pressures, considering power output reduced by the compressor power absorption. Differences of performance and efficiency between high and low pressure stacks have been noticed in the range of 10%. In operating at low pressure, higher efficiency is obtainable, but the maximum power of the stack is less; this means less fuel consumption, but requires a greater reacting surface and larger dimension of the stack. Consequently low pressures make the system simpler (a blower can be used instead of a compressor), and safer (there is practically no risk of breaking the membrane).

  4. Simulation and test of the thermal behavior of pressure switch

    Science.gov (United States)

    Liu, Yifang; Chen, Daner; Zhang, Yao; Dai, Tingting

    2018-04-01

    Little, lightweight, low-power microelectromechanical system (MEMS) pressure switches offer a good development prospect for small, ultra-long, simple atmosphere environments. In order to realize MEMS pressure switch, it is necessary to solve one of the key technologies such as thermal robust optimization. The finite element simulation software is used to analyze the thermal behavior of the pressure switch and the deformation law of the pressure switch film under different temperature. The thermal stress releasing schemes are studied by changing the structure of fixed form and changing the thickness of the substrate, respectively. Finally, the design of the glass substrate thickness of 2.5 mm is used to ensure that the maximum equivalent stress is reduced to a quarter of the original value, only 154 MPa when the structure is in extreme temperature (80∘C). The test results show that after the pressure switch is thermally optimized, the upper and lower electrodes can be reliably contacted to accommodate different operating temperature environments.

  5. [Non-animal toxicology in the safety testing of chemicals].

    Science.gov (United States)

    Heinonen, Tuula; Tähti, Hanna

    2013-01-01

    There is an urgent need to develop predictive test methods better than animal experiments for assessing the safety of chemical substances to man. According to today's vision this is achieved by using human cell based tissue and organ models. In the new testing strategy the toxic effects are assessed by the changes in the critical parameters of the cellular biochemical routes (AOP, adverse toxic outcome pathway-principle) in the target tissues. In vitro-tests are rapid and effective, and with them automation can be applied. The change in the testing paradigm is supported by all stakeholders: scientists, regulators and people concerned on animal welfare.

  6. Safety Test Report for the SNF Dry Storage System

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Seo, K. S.; Lee, J. H.; Lee, J. C.; Choi, W. S

    2008-11-15

    This is technical report conducted by KAERI under the contract with NETEC for safety test for the PWR S/F dry storage system. Leak Test was performed after drop test and turn-over test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the dry storage system is maintained. In the seismic test, the moving of the model was measured at SRTH seismic response of 0.4 g and 0.8 g. Therefore the seismic test results can be used fully to the test data for verification of the seismic analysis. In the thermal test, the direction of the inlet and outlet of the air has no effect on the heat transfer performance. The passive heat removal system of the horizontal storage module was designed well.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. Evaluation of micro fatigue crack growth under equi-biaxial stress by membranous pressure fatigue test

    International Nuclear Information System (INIS)

    Iida, Satoshi; Abe, Shigeki; Nakamura, Takao; Kamaya, Masayuki

    2014-01-01

    For preventing nuclear power plant (NPP) accidents, NPPs are required to ensure system safety in long term safe operation under aging degradation. Now, fatigue accumulation is one of major ageing phenomena and are evaluated to ensure safety by design fatigue curve that are based on the results of uniaxial fatigue tests. On the other hand, thermal stress that occurs in piping of actual components is not uniaxial but equi-biaxial. For accurate evaluation, it is required to conform real circumstance. In this study, membranous pressure fatigue test was conducted to simulated equi-biaxial stress. Crack initiation and crack growth were examined by replica investigation. Calculation result of equivalent stress intensity factor shows crack growth under equi-biaxial stress is faster than under uniaxial stress. It is concluded that equi-biaxial fatigue behavior should be considered in the evaluation of fatigue crack initiation and crack growth. (author)

  9. Heavy-Section Steel Technology Program intermediate-scale pressure vessel tests

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Smith, G.C.; Whitman, G.D.

    1977-01-01

    The tests of intermediate-size vessels with sharp flaws permitted the comparison of experimentally observed behavior with analytical predictions of the behavior of flawed pressure vessels. Fracture strains estimated by linear elastic fracture mechanics (LEFM) were accurate in the cases in which the flaws resided in regions of high transverse restraint and the fracture toughness was sufficiently low for unstable fracture to occur prior to yielding through the vessel wall. When both of these conditions were not present, unstable fracture did occur, always preceded by stable crack growth; and the cylinders with flaws initially less than halfway through the wall attained gross yield prior to burst. Predictions of failure pressure of the vessels with flawed nozzles, based upon LEFM estimates of failure strain, were very conservative. LEFM calculations of critical load were based upon small-specimen fracture toughness test data. Whenever gross yielding preceded failure, the actual strains achieved were considerably greater than the estimated strains at failure based on LEFM. In such cases the strength of the vessel may be no longer dependent upon plane-strain fracture toughness but upon the capacity of the cracked section to carry the imposed load stably in the plastic range. Stable crack growth, which has not been predictable quantitatively, is an important factor in elastic-plastic analysis of strength. The ability of the flawed vessels to attain gross yield in unflawed sections has important qualitative implications on pressure vessel safety margins. The gross yield condition occurs in light-water-reactor pressure vessels at about 2 x design pressure. The intermediate vessel tests that demonstrated a capacity for exceeding this load confirm that the presumed margin of safety is not diminished by the presence of flaws of substantial size, provided that material properties are adequate

  10. Experimental research on pressurized water reactor(PWR) safety

    International Nuclear Information System (INIS)

    Kim, Dong Su; Chae, Sung Ki; Chang, Won Pyo

    1991-12-01

    The objective of this research is to analyze the experimental results already performed in BETHSY facility of CEA France and to establish essential technologies for the future implementation of both such an experiment and computer code assessment, which are not undergoing in Korea so far. The contents of the present study are divided into 2 categories; namely, analysis of the BETHSY experimental data received from CEA, and CATHARE computer code simulation for the selected experiments, i.e. 'Natural Circulation(Test 4.3a)' and '2 Cold Leg Break'. The later studies are performed under the aims of CATHARE assessment as well as qualification of KOSAC code developing at KAERI, which is the subject in the next year and will concern an adequacy of KOSAC for the prediction of low flow natural circulation and a small break transients. (Author)

  11. Materials technology and the energy problem : application to the reliability and safety of nuclear pressure vessels

    International Nuclear Information System (INIS)

    Garrett, G.G.

    1975-01-01

    In the U.S.A. over the past few months, widespread plant shutdowns because of cracking problems has produced considerable public pressure for a reappraisal of the reliability and safety of nuclear reactors. The awareness of such problems, and their solution, is particularly relevant to South Africa at this time. Some materials problems related to nuclear plant failure are examined in this paper. Since catastrophic failure (without prior warning from slow leakage) is in principle possible for light water (pressurised) reactors under operating conditions, it is essential to maintain rigorous manufacturing and quality control procedures, in conjunction with thorough and frequent examination by non-destructive testing methods. Although tests currently in progress in the U.S.A. on large-scale model reactors suggest that mathematical stress and failure analyses, for simple geometries at least, are sound, current in situ surveillance programmes aimed at categorizing the effects of irradiation are inadequate. In addition, the effects on materials properties and subsequent fracture resistance of the combined effects of irradiation and thermal shock (arising from the injection of emergency cooling water during a loss-of coolant accident) are unknown. The problem of stress corrosion cracking in stainless steel pipelines is considerable, and at present virtually impossible to predict. Much of the available laboratory data is inapplicable in that it cannot account for the complex interactions of stress state, temperature, material variations and segregation effects, and water chemistry, especially in conjunction with irradiation effects, that are experienced in an operating environment

  12. Barometric pressure transient testing applications at the Nevada Test Site: formation permeability analysis. Final report

    International Nuclear Information System (INIS)

    Hanson, J.M.

    1984-12-01

    The report evaluates previous investigations of the gas permeability of the rock surrounding emplacement holes at the Nevada Test Site. The discussion sets the framework from which the present uncertainty in gas permeability can be overcome. The usefulness of the barometric pressure testing method has been established. Flow models were used to evaluate barometric pressure transients taken at NTS holes U2fe, U19ac and U20ai. 31 refs., 103 figs., 18 tabs

  13. Multilayer Pressure Vessel Materials Testing and Analysis. Phase 1

    Science.gov (United States)

    Cardinal, Joseph W.; Popelar, Carl F.; Page, Richard A.

    2014-01-01

    To provide NASA a comprehensive suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for aging multilayer pressure vessels, Southwest Research Institute (R) (SwRI) was contracted in two phases to obtain relevant material property data from a representative vessel. This report describes Phase 1 of this effort which includes a preliminary material property assessment as well as a fractographic, fracture mechanics and fatigue crack growth analyses of an induced flaw in the outer shell of a representative multilayer vessel that was subjected to cyclic pressure test. SwRI performed this Phase 1 effort under contract to the Digital Wave Corporation in support of their contract to Jacobs ATOM for the NASA Ames Research Center.

  14. Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7

    International Nuclear Information System (INIS)

    Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.; Bryan, R.H.

    1976-08-01

    The test of intermediate test vessel V-7 was a crack-initiation fracture test of a 152-mm-thick (6-in.), 990-mm-OD (39-in.) vessel of ASTM A533, grade B, class 1 steel plate with a sharp outside surface flaw 457 mm (18 in.) long and about 135 mm (5.3 in.) deep. The vessel was heated to 91 0 C (196 0 F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa

  15. 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

  16. Food safety issues of high pressure processed fruit/vegetable juices

    Czech Academy of Sciences Publication Activity Database

    Houška, M.; Strohalm, J.; Totušek, J.; Tříska, Jan; Vrchotová, Naděžda; Gabrovská, D.; Otová, B.; Gresová, P.

    2007-01-01

    Roč. 27, č. 1 (2007), s. 157-162 ISSN 0895-7959 R&D Projects: GA MZe QF3287 Institutional research plan: CEZ:AV0Z60870520 Keywords : Vegetable juices * High pressure processing * Food safety * Anti-mutagenic activity Subject RIV: GM - Food Processing Impact factor: 0.840, year: 2007

  17. A fracture mechanics and reliability based method to assess non-destructive testings for pressure vessels

    International Nuclear Information System (INIS)

    Kitagawa, Hideo; Hisada, Toshiaki

    1979-01-01

    Quantitative evaluation has not been made on the effects of carrying out preservice and in-service nondestructive tests for securing the soundness, safety and maintainability of pressure vessels, spending large expenses and labor. Especially the problems concerning the time and interval of in-service inspections lack the reasonable, quantitative evaluation method. In this paper, the problems of pressure vessels are treated by having developed the analysis method based on reliability technology and probability theory. The growth of surface cracks in pressure vessels was estimated, using the results of previous studies. The effects of nondestructive inspection on the defects in pressure vessels were evaluated, and the influences of many factors, such as plate thickness, stress, the accuracy of inspection and so on, on the effects of inspection, and the method of evaluating the inspections at unequal intervals were investigated. The analysis of reliability taking in-service inspection into consideration, the evaluation of in-service inspection and other affecting factors through the typical examples of analysis, and the review concerning the time of inspection are described. The method of analyzing the reliability of pressure vessels, considering the growth of defects and preservice and in-service nondestructive tests, was able to be systematized so as to be practically usable. (Kako, I.)

  18. Validation test case generation based on safety analysis ontology

    International Nuclear Information System (INIS)

    Fan, Chin-Feng; Wang, Wen-Shing

    2012-01-01

    Highlights: ► Current practice in validation test case generation for nuclear system is mainly ad hoc. ► This study designs a systematic approach to generate validation test cases from a Safety Analysis Report. ► It is based on a domain-specific ontology. ► Test coverage criteria have been defined and satisfied. ► A computerized toolset has been implemented to assist the proposed approach. - Abstract: Validation tests in the current nuclear industry practice are typically performed in an ad hoc fashion. This study presents a systematic and objective method of generating validation test cases from a Safety Analysis Report (SAR). A domain-specific ontology was designed and used to mark up a SAR; relevant information was then extracted from the marked-up document for use in automatically generating validation test cases that satisfy the proposed test coverage criteria; namely, single parameter coverage, use case coverage, abnormal condition coverage, and scenario coverage. The novelty of this technique is its systematic rather than ad hoc test case generation from a SAR to achieve high test coverage.

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  20. Materials, manufacture and testing of pressurized components of high-power steam power plants

    International Nuclear Information System (INIS)

    Blind, D.; Foehl, J.; Issler, L.; Schellhammer, W.; Sturm, D.; Kussmaul, K.; Heinrich, D.; Meyer, H.J.; Prestel, W.

    1981-01-01

    This is the first German review of materials, production and testing of pressure components of high-capacity steam power plants. The authors have been working in this field for years; their special subject has been the availability and reliability of pressure vessels, in particular in nuclear engineering. Fundamentals are presented as well as the findings obtained at the state Materials Testing Institute in Stuttgart. The material is presented in a well-structured classification; the most recent international findings, especially of the USA, are presented. This is possible due to the close cooperation between the Stuttgart institute and a number of US research institutes. The new subject of fracture mechanics is treated in some detail; its fundamentals are discussed from the American point of view while German considerations - in particular of the Reactor Safety Commission - are taken into account in the field of applications. (orig.) [de

  1. Test plan: Gas-threshold-pressure testing of the Salado Formation in the WIPP underground facility

    International Nuclear Information System (INIS)

    Saulnier, G.J. Jr.

    1992-03-01

    Performance assessment for the disposal of radioactive waste from the United States defense program in the WIPP underground facility must assess the role of post-closure was generation by waste degradation and the subsequent pressurization of the facility. be assimilated by the host formation will Whether or not the generated gas can be assimilated by the host formation will determine the ability of the gas to reach or exceed lithostatic pressure within the repository. The purpose of this test plan is (1) to present a test design to obtain realistic estimates of gas-threshold pressure for the Salado Formation WIPP underground facility including parts of the formation disturbed by the underground of the Salado, and (2) to provide a excavations and in the far-field or undisturbed part framework for changes and amendments to test objectives, practices, and procedures. Because in situ determinations of gas-threshold pressure in low-permeability media are not standard practice, the methods recommended in this testplan are adapted from permeability-testing and hydrofracture procedures. Therefore, as the gas-threshold-pressure testing program progresses, personnel assigned to the program and outside observers and reviewers will be asked for comments regarding the testing procedures. New and/or improved test procedures will be documented as amendments to this test plan, and subject to similar review procedures

  2. Ultrasonic testing of electron beam closure weld on pressure vessel

    International Nuclear Information System (INIS)

    Andrews, R.W.

    1975-01-01

    One of the special products manufactured at the General Electric Neutron Devices Department (GEND) is a small stainless steel vessel designed to hold a component under high pressure for long periods. The vessel is a thick-walled cylinder with a threaded receptacle into which a plug is screwed and welded after receiving the unit to be tested. The test cavity is then pressurized through a small diameter opening in the bottom and that opening is welded closed. When x-ray inspection techniques did not reveal defective welds at the threaded plug in a pressured vessel, occasional ''leakers'' occurred. With normal equipment tolerances, the electron beam spike tends to wander from the desired path, particularly at the root of the weld. Ultrasonic techniques were used to successfully inspect the weld. The testing technique is based on the observation that ultrasonic energy is reflected from the unwelded screw threads and not from the regions where the threads are completely fused together by welding. Any gas pore or any threaded region outside the weld bead can produce an echo. The units are rotated while the ultrasonic transducer travels in a direction parallel to the axis of rotation and toward the welded end. This produces a helical scan which is converted to a two-dimensional presentation in which incomplete welds can be noted. (U.S.)

  3. G-tunnel pressurized slot-testing evaluations

    International Nuclear Information System (INIS)

    Zimmerman, R.M.; Sifre-Soto, C.; Mann, K.L.; Bellman, R.A. Jr.; Luker, S.; Dodds, D.J.

    1992-04-01

    Designers and analysts of radioactive waste repositories must be able to predict the mechanical behavior of the host rock. Sandia National Laboratories elected to conduct a development program to enhance mechanical-type measurements. The program was focused on pressurized slot testing and featured (1) development of an improved method to cut slots using a chain saw with diamond-tipped cutters, (2) measurements useful for determining in situ stresses normal to slots, (3) measurements applicable for determining the in situ modulus of deformation parallel to a drift surface, and (4) evaluations of pressurized slot strength testing results and methods. This report contains data interpretation and evaluations. Included are recommendations for future efforts. This third report contains the interpretations of the testing with emphasis on the measurement results as they apply to describing rock behavior. In particular, emphases are placed on (1) normal stress determinations using the flatjack cancellation (FC) method, (2) modulus of deformation determinations, and (3) high pressure investigations. Most of the material in the first two reports is not repeated here. Appropriate data are repeated in tabular form

  4. High Pressure Hydrogen Pressure Relief Devices: Accelerated Life Testing and Application Best Practices

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, Robert M. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Post, Matthew B. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Buttner, William J. [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Rivkin, Carl H. [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-11-06

    Pressure relief devices (PRDs ) are used to protect high pressure systems from burst failure caused by overpressurization. Codes and standards require the use of PRDs for the safe design of many pressurized systems. These systems require high reliability due to the risks associated with a burst failure. Hydrogen service can increase the risk of PRD failure due to material property degradation caused by hydrogen attack. The National Renewable Energy Laboratory (NREL) has conducted an accelerated life test on a conventional spring loaded PRD. Based on previous failures in the field, the nozzles specific to these PRDs are of particular interest. A nozzle in a PRD is a small part that directs the flow of fluid toward the sealing surface to maintain the open state of the valve once the spring force is overcome. The nozzle in this specific PRD is subjected to the full tensile force of the fluid pressure. These nozzles are made from 440C material, which is a type of hardened steel that is commonly chosen for high pressure applications because of its high strength properties. In a hydrogen environment, however, 440C is considered a worst case material since hydrogen attack results in a loss of almost all ductility and thus 440C is prone to fatigue and material failure. Accordingly, 440C is not recommended for hydrogen service. Conducting an accelerated life test on a PRD with 440C material provides information on necessary and sufficient conditions required to produce crack initiation and failure. The accelerated life test also provides information on other PRD failure modes that are somewhat statistically random in nature.

  5. Recent progress in safety assessments of Japanese water cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Tsuru, Daigo; Enoeda, Mikio; Akiba, Masato

    2007-01-01

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 10 12 Bq, and tritium in purge gas is about 3 x 10 11 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  6. Work Pressure and Safety Behaviors among Health Workers in Ghana: The Moderating Role of Management Commitment to Safety

    Directory of Open Access Journals (Sweden)

    Kwesi Amponsah-Tawaih

    2016-12-01

    Conclusion: When employees perceive safety communication, safety systems and training to be positive, they seem to comply with safety rules and procedures than voluntarily participate in safety activities.

  7. Recommendations for safety testing with the in vivo comet assay.

    Science.gov (United States)

    Vasquez, Marie Z

    2012-08-30

    While the in vivo comet assay increases its role in regulatory safety testing, deliberations about the interpretation of comet data continue. Concerns can arise regarding comet assay publications with limited data from non-blind testing of positive control compounds and using protocols (e.g. dose concentrations, sample times, and tissues) known to give an expected effect. There may be a tendency towards bias when the validation or interpretation of comet assay data is based on results generated by widely accepted but non-validated assays. The greatest advantages of the comet assay are its sensitivity and its ability to detect genotoxicity in tissues and at sample times that could not previously be evaluated. Guidelines for its use and interpretation in safety testing should take these factors into account. Guidelines should be derived from objective review of data generated by blind testing of unknown compounds dosed at non-toxic concentrations and evaluated in a true safety-testing environment, where the experimental design and conclusions must be defensible. However, positive in vivo comet findings with such compounds are rarely submitted to regulatory agencies and this data is typically unavailable for publication due to its proprietary nature. To enhance the development of guidelines for safety testing with the comet assay, and with the permission of several sponsors, this paper presents and discusses relevant data from multiple GLP comet studies conducted blind, with unknown pharmaceuticals and consumer products. Based on these data and the lessons we have learned through the course of conducting these studies, I suggest significant adjustments to the current conventions, and I provide recommendations for interpreting in vivo comet assay results in situations where risk must be evaluated in the absence of carcinogenicity or clinical data. Copyright © 2012 Elsevier B.V. All rights reserved.

  8. Transient performance analysis of pressurized safety injection tank with a partition

    International Nuclear Information System (INIS)

    Bae, Youngmin; Kim, Young In; Kim, Keung Koo

    2015-01-01

    Highlights: • Functional performance of safety injection tanks with a partition is evaluated. • Effects of key design parameters are scrutinized. • Distinctive features of the flow in multi-unit safety injection tanks are explored. - Abstract: A parametric study has been performed to evaluate the functional performance of a pressurized multi-unit safety injection tank, which would be considered as one of the candidates for a passive safety injection system in a nuclear power plant. The influences of key design parameters including the orifice size, initial gas fraction, and resistance coefficients and operating condition on the injection flow rate are scrutinized with a discussion of the relevant flow features such as the choked flow of gas through an orifice and two interconnected regions of differing gaseous pressure. The obtained results indicate that a multi-unit safety injection tank can passively control the injection flow rate and provide a stable safety injection over a relatively long period even in the case of drastic depressurization of a reactor coolant system

  9. A study on the safety of TBP(150A) with forming analysis and strength test

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Kim, Hei Song

    2008-01-01

    For this study, the forming analyses and body strength tests of TBP were performed on the main pipe size 150A(KS D3507, KS D 3576 10S). The branched pipe sizes utilized were 25A, 32A, 40A, 50A, 65A, 80A, 100A and 125A. A general FEM program, ABAQUS, was used as the forming analyses method of TBP. Using the results, the strength of TBP was then tested in order to determine the safety of TBP when the working pressure was applied. The results indicate that TBP may be safely used in water-based fire protection pipe systems in terms of the strength

  10. Design and test of a high pressure centrifugal compressor

    International Nuclear Information System (INIS)

    Choi, Jae Ho; Han, Chak Heui; Paeng, Ki Seok; Chen, Seung Bae; Kim, Yong Ryun

    2005-01-01

    This paper presents an aerodynamic design, flow analysis and performance test of a pressure ratio 4:1 centrifugal compressor for gas turbine engine. The compressor is made up of a centrifugal impeller, a two-stage diffuser consisted of radial and axial types. The impeller has a 45 degree backswept angle and the design running tip clearance is 5% of impeller exit height. Three-dimensional numerical analysis is performed to analyze the flows in the impeller, diffuser and deswirler considering the impeller tip clearance. Test module and rig facilities for the compressor stage performance test are designed and fabricated. The overall compressor stage performances as well as the static pressure fields on the impeller and diffuser are measured. Two diffusers of wedge and airfoil types are tested with an impeller. The calculation and test results show that flow fields downstream the deswirler at the design and off-design points are highly nonuniform and the airfoil diffuser has the better aerodynamic characteristics than those of wedge diffuser

  11. Fracture Toughness Round Robin Test International in pressure tube materials

    International Nuclear Information System (INIS)

    Villagarcia, M.P.; Liendo, M.F.

    1993-01-01

    Part of the pressure tubes surveillance program of CANDU type reactors is to determine the fracture toughness using a special fracture specimen and test procedure. Atomic Energy of Canada Limited decided to hold a Round Robin Test International and 9 laboratories participated worldwide in which several pressure tube materials were selected: Zircaloy-2, Zr-2.5%Nb cold worked and Zr-2.5%Nb heat treated. The small specimens used held back the thickness and curvature of the tube. J-R curves at room temperature were obtained and the crack extension values were determined by electrical potential drop techniques. These values were compared with results generated from other laboratories and a bid scatter was founded. It could be due to slight variations in the test method or inhomogeneity of the materials and a statistical study must be done to see if there is any pattern. The next step for the Round Robin Test would be to make some modifications in the test method in order to reduce the scatter. (Author)

  12. Large scale steam flow test: Pressure drop data and calculated pressure loss coefficients

    International Nuclear Information System (INIS)

    Meadows, J.B.; Spears, J.R.; Feder, A.R.; Moore, B.P.; Young, C.E.

    1993-12-01

    This report presents the result of large scale steam flow testing, 3 million to 7 million lbs/hr., conducted at approximate steam qualities of 25, 45, 70 and 100 percent (dry, saturated). It is concluded from the test data that reasonable estimates of piping component pressure loss coefficients for single phase flow in complex piping geometries can be calculated using available engineering literature. This includes the effects of nearby upstream and downstream components, compressibility, and internal obstructions, such as splitters, and ladder rungs on individual piping components. Despite expected uncertainties in the data resulting from the complexity of the piping geometry and two-phase flow, the test data support the conclusion that the predicted dry steam K-factors are accurate and provide useful insight into the effect of entrained liquid on the flow resistance. The K-factors calculated from the wet steam test data were compared to two-phase K-factors based on the Martinelli-Nelson pressure drop correlations. This comparison supports the concept of a two-phase multiplier for estimating the resistance of piping with liquid entrained into the flow. The test data in general appears to be reasonably consistent with the shape of a curve based on the Martinelli-Nelson correlation over the tested range of steam quality

  13. Demonstration tests for low level radioactive waste packaging safety

    International Nuclear Information System (INIS)

    Nagano, I.; Shimura, S.; Miki, T.; Tamamura, T.; Kunitomi, K.

    1993-01-01

    The transport packaging for low level radioactive waste (so-called the LLW packaging) has been developed to be utilized for transportation of LLW in 200 liter-drums from Japanese nuclear power stations to the LLW Disposal Center at Rokkashomura in Aomori Prefecture. Transportation is expected to start from December in 1992. We will explain the brief history of the development, technical features and specifications as well as two kinds of safety demonstration tests, namely one is '1.2 meter free drop test' and the other is 'ISO container standard test'. (J.P.N.)

  14. New methods for the safety testing of transgenic food

    DEFF Research Database (Denmark)

    Knudsen, Ib; Poulsen, Morten; Kledal, S. T.

    2004-01-01

    for guiding the precise design of the animal study. The genetically modified food plants to be used for this test development will be 3 transgenic rice varieties (2 types of lectins and the Bt toxin). Objectives The overall objective of this project is to develop and validate the scientific methodology which......Background This project proposal deals with the development of a sensitive and specific animal test which is necessary for safety analysis of genetically modified plants according to the Opinion of the Scientific Committee for Food on the assessment of novels foods. The test will be based...

  15. Posttest analysis of the FFTF inherent safety tests

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Claybrook, S.W.

    1987-01-01

    Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactor and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core and plant analysis was not completed in time to perform pretest analyses. The purpose of this paper is to present the results of the posttest analysis of the 1986 FFTF inherent safety tests using the SASSYS code

  16. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    International Nuclear Information System (INIS)

    Wright, A.E.; Dutt, D.S.; Harrison, L.J.

    1990-01-01

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

  17. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  18. Sealing performance test for main flange of pressure vessel of T2 test section in HENDEL

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Inagaki, Yoshiyuki; Matsumoto, Kiminori; Kondou, Yasuo; Suzuki, Kunihiko; Miyamoto, Yoshiaki; Asami, Masanobu.

    1990-12-01

    A pressure vessel of T 2 test section in helium engineering demonstration loop (HENDEL) was fabricated to the same scale of the reactor pressure vessel made of 2(1/4)Cr-1Mo steel in high temperature engineering test reactor (HTTR). Also, the sealing structure of a main flange of pressure vessel in T 2 test section was composed of the double metal O-rings and Ω-seal which would be used in the sealing structure of HTTR. The sealing performance test for the main flange of the pressure vessel in T 2 test section was carried out to confirm the integrity of sealing structure of a main flange in HTTR. T 2 test section has been operated about 7700 hours in previous 18 cycles. The leakage of helium gas from inner metal O-ring was measured by the static pressurized process under the operating condition of HTTR (helium gas: 400degC, 40kg/cm 2 G, 4gk/s). The calculated leakage of helium gas was less than 9.6x10 -7 atm·cm 3 /sec. From the result, it is expected that the sealing structure of main flange in HTTR would maintain the leak tightness in the life. (author)

  19. Present status of the disk pressure tests for hydrogen embrittlement

    International Nuclear Information System (INIS)

    Fidelle, J.P.

    1985-05-01

    The Disk Pressure Tests (DPT) have been developed considerably theoretically and experimentally for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for Environment Embrittlement due to H 2 , hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably for pressure up to 300 MPa and temperature (-160 0 C to 1000 0 C). Very low strain rate -longer than a month- tests have been able to evidence embrittlement of FFC alloys where H diffusivity is low. Conversely for very oxidation - sensitive metals (e.g. Nb and Ta) effects may appear only at somewhat high rates. The relationship between dynamic (increasing stress) tests, static (delayed failure) and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analyzed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 150 + materials in different conditions. From the tests on a large number of metal systems, a theory of HE has been derived which accounts for the behavior of metals and alloys either embrittled and or hydrited. Finally comparison of HGE tests and service behavior of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service

  20. Safety test facilities. Needs and concepts. A French evaluation

    International Nuclear Information System (INIS)

    Tretiakoff, O.; Bailly, J.

    1976-01-01

    The fuel behaviour of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program (local blockage, sudden flow reduction and pump coast-down at constant power). These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The paper explains how a logical set of simple observations has led to the present state of the Cadarache in-pile experimental safety program and how it may help to find the way in a dense forest of both technical and psychological difficulties

  1. Safety test facilities. Needs and concepts. A French evaluation

    International Nuclear Information System (INIS)

    Tretiakoff, O.; Bailly, J.

    1976-01-01

    The fuel behavior of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program. These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The purpose of this paper is to explain how a logical set of simple observations has led us to the present state of the Cadarache in-pile experimental safety program and how it may help us to find our way in a dense forest of both technical and psychological difficulties

  2. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  3. Pressure and pressure derivative analysis for injection tests with variable temperature without type-curve matching

    International Nuclear Information System (INIS)

    Escobar, Freddy Humberto; Martinez, Javier Andres; Montealegre Matilde

    2008-01-01

    The analysis of injection tests under nonisothermic conditions is important for the accurate estimation of the reservoir permeability and the well's skin factor; since previously an isothermical system was assumed without taking into account a moving temperature front which expands with time plus the consequent changes in both viscosity and mobility between the cold and the hot zone of the reservoir which leads to unreliable estimation of the reservoir and well parameters. To construct the solution an analytical approach presented by Boughrara and Peres (2007) was used. That solution was initially introduced for the calculation of the injection pressure in an isothermic system. It was later modified by Boughrara and Reynolds (2007) to consider a system with variable temperature in vertical wells. In this work, the pressure response was obtained by numerical solution of the anisothermical model using the Gauss Quadrature method to solve the integrals, and assuming that both injection and reservoir temperatures were kept constant during the injection process and the water saturation is uniform throughout the reservoir. For interpretation purposes, a technique based upon the unique features of the pressure and pressure derivative curves were used without employing type-curve matching (TDS technique). The formulation was verified by its application to field and synthetic examples. As expected, increasing reservoir temperature causes a decrement in the mobility ratio, then estimation of reservoir permeability is some less accurate from the second radial flow, especially, as the mobility ratio increases

  4. Testing laboratories, its function in ensuring industrial safety

    International Nuclear Information System (INIS)

    Sanchez Fernandez, M.

    2015-01-01

    This article discusses and justifies the development of industrial laboratories (testing and calibration) in Spain, since its embryo, its creation and development, to the present day. Likewise, presents its interrelation with other agents, as well as the legislative and technical framework is application along to the years. Within this development of the sector, highlights the period of the conformity assessment, and consequently its relationship with Industrial safety. Finally, describes the organizational situation of the sector both nationally and internationally. (Author)

  5. The electron test accelerator safety in design and operation

    International Nuclear Information System (INIS)

    McKeown, J.

    1980-06-01

    The Electron Test Accelerator is being designed as an experiment in accelerator physics and technology. With an electron beam power of up to 200 kW the operation of the accelerator presents a severe radiation hazard as well as rf and electrical hazards. The design of the safety system provides fail-safe protection while permitting flexibility in the mode of operation and minimizing administrative controls. (auth)

  6. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  7. Evaluation of seven in vitro alternatives for ocular safety testing.

    Science.gov (United States)

    Bruner, L H; Kain, D J; Roberts, D A; Parker, R D

    1991-07-01

    Seven in vitro assays were evaluated to determine if any were useful as screening procedures in ocular safety assessment. Seventeen test materials (chemicals, household cleaners, hand soaps, dishwashing liquids, shampoos, and liquid laundry detergents) were tested in each assay. In vivo ocular irritation scores for the materials were obtained from existing rabbit low volume eye test (LVET) data. The seven assays evaluated included the silicon microphysiometer (SM), luminescent bacteria toxicity test (LBT), neutral red assay (NR), total protein assay (TP), Tetrahymena thermophila motility assay (TTMA), bovine eye/chorioallantoic membrane assay (BE/CAM), and the EYTEX system (ETS). For the seventeen materials used in this study there was a significant correlation between the in vivo irritant potential and in vitro data for all the tests except the EYTEX System (SM, r = -0.87; LBT, r = -0.91; NR, r = -0.85; TTMA, r = 0.78; TP, r = -0.86; ETS, r = 0.29). The irritation classifications provided by the BE/CAM also did not correspond with the actual in vivo irritancy potential of the test materials. The result of this study suggested it may be possible to classify materials into broad irritancy categories with some of the assays. This would allow their use as screens prior to limited in vivo confirmation in the ocular safety assessment process.

  8. Using partial safety factors in wind turbine design and testing

    Energy Technology Data Exchange (ETDEWEB)

    Musial, W.D. [National Renewable Energy Lab., Golden, CO (United States)

    1997-12-31

    This paper describes the relationship between wind turbine design and testing in terms of the certification process. An overview of the current status of international certification is given along with a description of limit-state design basics. Wind turbine rotor blades are used to illustrate the principles discussed. These concepts are related to both International Electrotechnical Commission and Germanischer Lloyd design standards, and are covered using schematic representations of statistical load and material strength distributions. Wherever possible, interpretations of the partial safety factors are given with descriptions of their intended meaning. Under some circumstances, the authors` interpretations may be subjective. Next, the test-load factors are described in concept and then related to the design factors. Using technical arguments, it is shown that some of the design factors for both load and materials must be used in the test loading, but some should not be used. In addition, some test factors not used in the design may be necessary for an accurate test of the design. The results show that if the design assumptions do not clearly state the effects and uncertainties that are covered by the design`s partial safety factors, outside parties such as test labs or certification agencies could impose their own meaning on these factors.

  9. Results of the 1986 FFTF inherent safety tests

    International Nuclear Information System (INIS)

    Burke, T.M.; Campbell, L.R.; Franz, G.R.; Knecht, W.L.

    1987-01-01

    A series of tests was recently completed at the 400-MW (thermal) Fast Flux Test Facility (FFTF) to further demonstrate the passive safety characteristics of liquid-metal-cooled fast reactors. Earlier FFTF testing of decay heat removal by sodium natural circulation was reported in 1981. The main purpose of the 1986 test series was to demonstrate passive reactor shutdown during a loss-of-flow event when several inherent shutdown devices called gas expansion modules (GEMs) were installed in the reactor. However, these tests also provide further data on the natural circulation performance of the primary system, in particular the reactor core, and thus add to the data base available for checking the validity of available analytical tools

  10. Active earth pressure model tests versus finite element analysis

    Science.gov (United States)

    Pietrzak, Magdalena

    2017-06-01

    The purpose of the paper is to compare failure mechanisms observed in small scale model tests on granular sample in active state, and simulated by finite element method (FEM) using Plaxis 2D software. Small scale model tests were performed on rectangular granular sample retained by a rigid wall. Deformation of the sample resulted from simple wall translation in the direction `from the soil" (active earth pressure state. Simple Coulomb-Mohr model for soil can be helpful in interpreting experimental findings in case of granular materials. It was found that the general alignment of strain localization pattern (failure mechanism) may belong to macro scale features and be dominated by a test boundary conditions rather than the nature of the granular sample.

  11. Testing of a portable ultrahigh pressure water decontamination system (UHPWDS)

    International Nuclear Information System (INIS)

    Lovell, A.; Dahlby, J.

    1996-02-01

    This report describes the tests done with a portable ultrahigh pressure water decontamination system (UHPWDS) on highly radioactively contaminated surfaces. A small unit was purchased, modified, and used for in-situ decontamination to change the waste level of the contaminated box from transuranic (TRU) waste to low- level waste (LLW). Low-level waste is less costly by as much as a factor of five or more if compared with TRU waste when handling, storage, and disposal are considered. The portable unit we tested is commercially available and requires minimal utilities for operation. We describe the UHPWDS unit itself, a procedure for its use, the results of the testing we did, and conclusions including positive and negative aspects of the UHPWDS

  12. A modified split Hopkinson pressure bar for toughness tests

    Science.gov (United States)

    Granier, N.; Grunenwald, T.

    2006-08-01

    In order to characterize material toughness or to study crack arrest under dynamic loading conditions, a new testing device has been developed at CEA/Valduc. A new Split Hopkinson Pressure Bar (SHPB) has been modified: it is now composed of a single incident bar and a double transmitter bar. With this facility, a notched specimen can be loaded under three points bending conditions. Qualification tests with titanium and steel notched samples are presented. Data treatment software has been adapted to estimate the sample deflection as a function of time and treat the energy balance. These results are compared with classical Charpy experiments. Effect of various contact areas between specimen and bars are studied to point out their influence on obtained measurements. The advantage of a “knife” contact compared to a plane one is then clearly demonstrated. All results obtained with this new testing device are in good agreement and show a reduced scattering.

  13. Hydraulic testing in granite using the sinusoidal variation of pressure

    International Nuclear Information System (INIS)

    Black, J.H.; Holmes, D.C.; Noy, D.J.

    1982-09-01

    Access to two boreholes at the Carwynnen test site in Cornwall enabled the trial of a number of innovative approaches to the hydrogeology of fractured crystalline rock. These methods ranged from the use of seisviewer data to measure the orientation of fractures to the use of the sinusoidal pressure technique to measure directional hydraulic diffusivity. The testing began with a short programme of site investigation consisting of borehole caliper and seisviewer logging followed by some single borehole hydraulic tests. The single borehole hydraulic testing was designed to assess whether the available boreholes and adjacent rock were suitable for testing using the sinusoidal method. The main testing methods were slug and pulse tests and were analysed using the fissured porous medium analysis proposed in Barker and Black (1983). Derived hydraulic conductivity (K) ranged from 2 x 10 -12 m/sec to 5 x 10 -7 m/sec with one near-surface zone of high K being perceived in both boreholes. The results were of the form which is typical of fractured rock and indicated a combination of high fracture frequency and permeable granite matrix. The results are described and discussed. (author)

  14. Unavoidable Pressure Ulcers: Development and Testing of the Indiana University Health Pressure Ulcer Prevention Inventory.

    Science.gov (United States)

    Pittman, Joyce; Beeson, Terrie; Terry, Colin; Dillon, Jill; Hampton, Charity; Kerley, Denise; Mosier, Judith; Gumiela, Ellen; Tucker, Jessica

    2016-01-01

    Despite prevention strategies, hospital-acquired pressure ulcers (HAPUs) continue to occur in the acute care setting. The purpose of this study was to develop an operational definition of and an instrument for identifying avoidable/unavoidable HAPUs in the acute care setting. The Indiana University Health Pressure Ulcer Prevention Inventory (PUPI) was developed and psychometric testing was performed. A retrospective pilot study of 31 adult hospitalized patients with an HAPU was conducted using the PUPI. Overall content validity index of 0.99 and individual item content validity index scores (0.9-1.0) demonstrated excellent content validity. Acceptable PUPI criterion validity was demonstrated with no statistically significant differences between wound specialists' and other panel experts' scoring. Construct validity findings were acceptable with no statistically significant differences among avoidable or unavoidable HAPU patients and their Braden Scale total scores. Interrater reliability was acceptable with perfect agreement on the total PUPI score between raters (κ = 1.0; P = .025). Raters were in total agreement 93% (242/260) of the time on all 12 individual PUPI items. No risk factors were found to be significantly associated with unavoidable HAPUs. An operational definition of and an instrument for identifying avoidable/unavoidable HAPUs in the acute care setting were developed and tested. The instrument provides an objective and structured method for identifying avoidable/unavoidable HAPUs. The PUPI provides an additional method that could be used in root-cause analyses and when reporting adverse pressure ulcer events.

  15. Technology Solutions Case Study: Combustion Safety Simplified Test Protocol

    Energy Technology Data Exchange (ETDEWEB)

    L. Brand, D. Cautley, D. Bohac, P. Francisco, L. Shen, and S. Gloss

    2015-12-01

    Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives.

  16. 46 CFR 61.30-20 - Automatic control and safety tests.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Automatic control and safety tests. 61.30-20 Section 61... TESTS AND INSPECTIONS Tests and Inspections of Fired Thermal Fluid Heaters § 61.30-20 Automatic control and safety tests. Operational tests and checks of all safety and limit controls, combustion controls...

  17. Temperature and pressure instrumentation in WWERs and their testing

    International Nuclear Information System (INIS)

    Por, G.

    1998-01-01

    A description of WWER model V-213 reactors of second generation is presented and compared to analogous NPPs including description of temperature and pressure instrumentation which was tested at Paks NPP. From the experimental results it was concluded that measured response of in core neutron detector to bubbles strongly depends on the relative position of detector and point bubble injection. Neutron noise spectra show characteristic sink when the origin of bubbles is close to the detectors. Dependence of phase behaviour on the boiling conditions is included as well

  18. NEK containment integrated leak rate test at full pressure

    International Nuclear Information System (INIS)

    Skaler, F.; Planinc, V.; Gregoric, D.; Cicvaric, D.

    1999-01-01

    NPP Krsko is a Pressure Water Reactor (PWR) Plant which has four barriers to prevent release of radioactive fission products. These four barriers are following: Fuel itself, Fuel Clad, Reactor Coolant System and Containment Building. Containment is the last barrier which can prevent release of fission product when other barriers have been already broken. To find out the real condition of containment vessel and to prove its ability of withstanding increased parameters during accident we have to perform Containment Integrated Leak Rate Test at least three times in every ten years of operation. CILRT 1999 in NPP Krsko was completely performed following regulation of 10CFR50 App. J Option A and ANSI/ANS 56.8-1987. The main goal of CILRT is to prove that the leakage of containment pathways and wall structures are within limits prescribed in Technical Specifications by pressurization of containment building above peak accident pressure Pa and measuring the mass changes of air using Ideal Gas Law.(author)

  19. Steam Pressure-Reducing Station Safety and Energy Efficiency Improvement Project

    Energy Technology Data Exchange (ETDEWEB)

    Lower, Mark D [ORNL; Christopher, Timothy W [ORNL; Oland, C Barry [ORNL

    2011-06-01

    The Facilities and Operations (F&O) Directorate is sponsoring a continuous process improvement (CPI) program. Its purpose is to stimulate, promote, and sustain a culture of improvement throughout all levels of the organization. The CPI program ensures that a scientific and repeatable process exists for improving the delivery of F&O products and services in support of Oak Ridge National Laboratory (ORNL) Management Systems. Strategic objectives of the CPI program include achieving excellence in laboratory operations in the areas of safety, health, and the environment. Identifying and promoting opportunities for achieving the following critical outcomes are important business goals of the CPI program: improved safety performance; process focused on consumer needs; modern and secure campus; flexibility to respond to changing laboratory needs; bench strength for the future; and elimination of legacy issues. The Steam Pressure-Reducing Station (SPRS) Safety and Energy Efficiency Improvement Project, which is under the CPI program, focuses on maintaining and upgrading SPRSs that are part of the ORNL steam distribution network. This steam pipe network transports steam produced at the ORNL steam plant to many buildings in the main campus site. The SPRS Safety and Energy Efficiency Improvement Project promotes excellence in laboratory operations by (1) improving personnel safety, (2) decreasing fuel consumption through improved steam system energy efficiency, and (3) achieving compliance with applicable worker health and safety requirements. The SPRS Safety and Energy Efficiency Improvement Project being performed by F&O is helping ORNL improve both energy efficiency and worker safety by modifying, maintaining, and repairing SPRSs. Since work began in 2006, numerous energy-wasting steam leaks have been eliminated, heat losses from uninsulated steam pipe surfaces have been reduced, and deficient pressure retaining components have been replaced. These improvements helped ORNL

  20. Cyber Security Test Strategy for Non-safety Display System

    International Nuclear Information System (INIS)

    Son, Han Seong; Kim, Hee Eun

    2016-01-01

    Cyber security has been a big issue since the instrumentation and control (I and C) system of nuclear power plant (NPP) is digitalized. A cyber-attack on NPP should be dealt with seriously because it might cause not only economic loss but also the radioactive material release. Researches on the consequences of cyber-attack onto NPP from a safety point of view have been conducted. A previous study shows the risk effect brought by initiation of event and deterioration of mitigation function by cyber terror. Although this study made conservative assumptions and simplifications, it gives an insight on the effect of cyber-attack. Another study shows that the error on a non-safety display system could cause wrong actions of operators. According to this previous study, the failure of the operator action caused by a cyber-attack on a display system might threaten the safety of the NPP by limiting appropriate mitigation actions. This study suggests a test strategy focusing on the cyber-attack on the information and display system, which might cause the failure of operator. The test strategy can be suggested to evaluate and complement security measures. Identifying whether a cyber-attack on the information and display system can affect the mitigation actions of operator, the strategy to obtain test scenarios is suggested. The failure of mitigation scenario is identified first. Then, for the test target in the scenario, software failure modes are applied to identify realistic failure scenarios. Testing should be performed for those scenarios to confirm the integrity of data and to assure effectiveness of security measures

  1. Cyber Security Test Strategy for Non-safety Display System

    Energy Technology Data Exchange (ETDEWEB)

    Son, Han Seong [Joongbu University, Geumsan (Korea, Republic of); Kim, Hee Eun [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    Cyber security has been a big issue since the instrumentation and control (I and C) system of nuclear power plant (NPP) is digitalized. A cyber-attack on NPP should be dealt with seriously because it might cause not only economic loss but also the radioactive material release. Researches on the consequences of cyber-attack onto NPP from a safety point of view have been conducted. A previous study shows the risk effect brought by initiation of event and deterioration of mitigation function by cyber terror. Although this study made conservative assumptions and simplifications, it gives an insight on the effect of cyber-attack. Another study shows that the error on a non-safety display system could cause wrong actions of operators. According to this previous study, the failure of the operator action caused by a cyber-attack on a display system might threaten the safety of the NPP by limiting appropriate mitigation actions. This study suggests a test strategy focusing on the cyber-attack on the information and display system, which might cause the failure of operator. The test strategy can be suggested to evaluate and complement security measures. Identifying whether a cyber-attack on the information and display system can affect the mitigation actions of operator, the strategy to obtain test scenarios is suggested. The failure of mitigation scenario is identified first. Then, for the test target in the scenario, software failure modes are applied to identify realistic failure scenarios. Testing should be performed for those scenarios to confirm the integrity of data and to assure effectiveness of security measures.

  2. PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

    International Nuclear Information System (INIS)

    Paladino, D.; Dreier, J.

    2012-01-01

    The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

  3. THYC qualification on Vatican-1 low pressure tests

    International Nuclear Information System (INIS)

    Duval, C.; Guichard, J.

    1991-06-01

    PWR cores or fuel assemblies are components of a nuclear power plant involving single and two-phase flows in rod bundles. The knowledge of the detailed two-phase and three-dimensional flow patterns is necessary to evaluate the singularity (grids) and bypass effects on the Departure from Nucleate Boiling (DNB) in reactor cores during incidental transients. For that purpose, since 1989, the VATICAN experiment has been performed at EDF as a part of the qualification program of the three-dimensional computer code THYC, developed by EDF. The qualification strategy of the THYC software for PWR cores is the following: assuming the theoretical or experimental knowledge of regular and singular pressure drops and grid turbulence sources in single-phase, pressure drop multipliers and relative velocity in two-phase flow, the VATICAN experiment allows to evaluate the diffusion phenomena in two-phase flow. It provides thermalhydraulic measurements on a mock-up of a part of 900 MWe PWR fuel assembly in single and two-phase flows, with power and quality gradients. The first configuration of the mock-up, with simple spacer grids, is studied (VATICAN-1). The specific effects of mixing spacer grids will be compared to these data through a second configuration. The last void fraction measurements, using a γ-ray technique, performed on VATICAN-1 low pressure tests allowed to qualify a set of closure relations, particularly a model of little two-phase diffusion, adapted to two-phase flows at low pressure (5.0MPa). The qualification of subcooled boiling and diffusion models will continue on next VATICAN and other experimental campaigns [fr

  4. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  5. Safety of 5 MW district heating reactor (DHR) and hydraulic dynamic pressure drive control rods

    International Nuclear Information System (INIS)

    Wu Yuanqiang; Wang Dazhong

    1991-11-01

    The principles and movement characteristic of the hydraulic dynamic pressure drive for control rods in 5 MW district heating reactor are described with stress on analysis of its effects on reactor safety features. The drive is different from electric-magnetic drive for PWR or hydraulic drive for BWR. The drive cylinder is driven by dynamic pressure. In the new drive system, the reactor coolant (water) used as actuating medium is pressed by pump, then injected into a step cylinder which is set in the reactor core. The cylinder will move step by step by controlling flow, then the cylinder drives the neutron absorber and controls nuclear reaction. The drive is characterized by simplicity in structure, high reliability, inherent safety, reduction in reactor height, economy, etc

  6. Safety system function trend indicator: Theory and test application

    International Nuclear Information System (INIS)

    Azarm, M.A.; Carbonaro, J.F.; Boccio, J.L.; Vesely, W.E.

    1989-01-01

    The purpose of this paper is to summarize research conducted on the development and validation of quantitative indicators of safety performance. This work, performed under the Risk-Based Performance Indicator (RBPI) Project, FIN A-3295, for the Office of Research (RES), is considered part of NRC's Performance Indicator Program which is being coordinated through the Office for the Analysis and Evaluation of Operational Data (AEOD). The program originally focused on risk-based indicators at high levels of safety indices (e.g., core-damage frequency, functional unavailabilities, and sequence monitoring). The program was then redirected towards a more amenable goal, safety system unavailability indicators, mainly due to the lack of PRA models and plant data. In that regard, BNL published a technical report that introduced the concept of cycle-based indicators and also described various alternatives of monitoring safety system unavailabilities. Further simplification of these indicators was requested by NRC to facilitate their applications to all plants in a timely manner. This resulted in the development of Safety System Function Trend (SSFT) indicators which minimize the need for detailed system model as well as component history. The theoretical bases for these indicators were developed through various simulation studies to determine the ease of detecting a trend and/or unacceptable performance. These indicators, along with several other indicators, were then generated and compared using plant data as a part of a test application. The SSFT indicators, specifically, were constructed for a total of eight plants, consisting of two systems per plant. Emphasis was placed on examining relative changes, as well as the indicator's actual level. Both the trend and actual indicator level were found to be important in identifying plants with potential problems

  7. Safety assessment of pipes with multiple local wall thinning defects under pressure and bending moment

    International Nuclear Information System (INIS)

    Peng Jian; Zhou Changyu; Xue Jilin; Dai Qiao; He Xiaohua

    2011-01-01

    The safety assessment of pipes with local wall thinning defects is highly important in engineering. Most attention has been paid on the safety assessment of pipe with single local wall thinning defect, while the studies about multiple local wall thinning defects are not nearly enough. However, the interaction of multiple local wall thinning defects in some conditions is great, and may have a great impact on the safety assessment. In the present standard API 579/ASME FFS, the safety assessment of pipes with multiple local wall thinning defects is given, while as well as the influence of load condition, the influences of arrangement and relative depth of defects are ignored, which may influence the safety assessment considerably. In this paper, the influence of the interaction between multiple local wall thinning defects on the remaining strength of pipes at different arrangements and depths of defects under different load conditions (pressure, tension-bending moment and compression-bending moment) are studied. A quantified index is defined to describe the interaction between defects quantitatively. For different arrangements and relative depths of defects, based on a limit value 0.05 of the quantified index of the interaction between defects, a relatively systematic safety assessment of pipes with multiple local wall thinning defects under different load conditions has been proposed.

  8. Study of the suit inflation effect on crew safety during landing using a full-pressure IVA suit for new-generation reentry space vehicles

    Science.gov (United States)

    Wataru, Suzuki

    Recently, manned space capsules have been recognized as beneficial and reasonable human space vehicles again. The Dragon capsule already achieved several significant successes. The Orion capsule is going to be sent to a high-apogee orbit without crews for experimental purposes in September 2014. For such human-rated space capsules, the study of acceleration impacts against the human body during splashdown is essential to ensure the safety of crews. Moreover, it is also known that wearing a full pressure rescue suit significantly increases safety of a crew, compared to wearing a partial pressure suit. This is mainly because it enables the use of a personal life support system independently in addition to that which installed in the space vehicle. However, it is unclear how the inflation of the full pressure suit due to pressurization affects the crew safety during splashdown, especially in the case of the new generation manned space vehicles. Therefore, the purpose of this work is to investigate the effect of the suit inflation on crew safety against acceleration impact during splashdown. For this objective, the displacements of the safety harness in relation with the suit, a human surrogate, and the crew seats during pressurizing the suit in order to determine if the safety and survivability of a crew can be improved by wearing a full pressure suit. For these tests, the DL/H-1 full pressure IVA suit, developed by Pablo de Leon and Gary L. Harris, will be used. These tests use image analysis techniques to determine the displacements. It is expected, as a result of these tests, that wearing a full pressure suit will help to mitigate the impacts and will increase the safety and survivability of a crew during landing since it works as a buffer to mitigate impact forces during splashdown. This work also proposes a future plan for sled test experiments using a sled facility such as the one in use by the Civil Aerospace Medical Institute (CAMI) for experimental validation

  9. Pressurized helium II-cooled magnet test facility

    International Nuclear Information System (INIS)

    Warren, R.P.; Lambertson, G.R.; Gilbert, W.S.; Meuser, R.B.; Caspi, S.; Schafer, R.V.

    1980-06-01

    A facility for testing superconducting magnets in a pressurized bath of helium II has been constructed and operated. The cryostat accepts magnets up to 0.32 m diameter and 1.32 m length with current to 3000 A. In initial tests, the volume of helium II surrounding the superconducting magnet was 90 liters. Minimum temperature reached was 1.7 K at which point the pumping system was throttled to maintain steady temperature. Helium II reservoir temperatures were easily controlled as long as the temperature upstream of the JT valve remained above T lambda; at lower temperatures control became difficult. Positive control of the temperature difference between the liquid and cold sink by means of an internal heat source appears necessary to avoid this problem. The epoxy-sealed vessel closures, with which we have had considerable experience with normal helium vacuum, also worked well in the helium II/vacuum environment

  10. Applications of High and Ultra High Pressure Homogenization for Food Safety

    OpenAIRE

    Patrignani, Francesca; Lanciotti, Rosalba

    2016-01-01

    Traditionally, the shelf-life and safety of foods have been achieved by thermal processing. Low temperature long time (LTLT) and high temperature short time (HTST) treatments are the most commonly used hurdles for the pasteurization of fluid foods and raw materials. However, the thermal treatments can reduce the product quality and freshness. Consequently, some non-thermal pasteurization process have been proposed during the last decades, including high hydrostatic pressure (HHP), pulsed ele...

  11. RELAP5 - a new tool for pressurized water reactor safety analysis

    International Nuclear Information System (INIS)

    Perneczky, L.

    1988-11-01

    The RELAP type pressurized water reactor safety system codes are used world wide for the loss of coolant accident analyses. In this paper the RELAP5, the advanced generation of the code family is presented. The relationship to RELAP4/mod6 version is discussed. The capability of the RELAP5/mod1-EUR version for small, medium and large break LOCA is investigated based on international user experience. (author) 30 refs

  12. Radiation pressure calibration and test mass reflectivities for LISA Pathfinder

    International Nuclear Information System (INIS)

    Korsakova, Natalia; Kaune, Brigitte

    2017-01-01

    This paper describes a series of experiments which were carried out during the main operations of LISA Pathfinder. These experiments were performed by modulating the power of the measurement and reference beams. In one series of experiments the beams were sequentially switched on and off. In the other series of experiments the powers of the beams were modulated within 0.1% and 1% of the constant power. These experiments use recordings of the total power measured on the photodiodes to infer the properties of the Optical Metrology System (OMS), such as reflectivities of the test masses and change of the photodiode efficiencies with time. In the first case the powers are back propagated from the different photodiodes to the same place on the optical bench to express the unknown quantities in the measurement with the complimentary photodiode measurements. They are combined in the way that the only unknown left is the test mass reflectivities. The second experiment compared two estimates of the force applied to the test masses due to the radiation pressure that appears because of the beam modulations. One estimate of the force is inferred from the measurements of the powers on the photodiodes and propagation of this measurement to the test masses. The other estimation of the force is done by calculating it from the change in the main scientific output of the instrument – differential displacement of the two test masses. (paper)

  13. Radiation pressure calibration and test mass reflectivities for LISA Pathfinder

    Science.gov (United States)

    Korsakova, Natalia; Kaune, Brigitte; LPF Collaboration

    2017-05-01

    This paper describes a series of experiments which were carried out during the main operations of LISA Pathfinder. These experiments were performed by modulating the power of the measurement and reference beams. In one series of experiments the beams were sequentially switched on and off. In the other series of experiments the powers of the beams were modulated within 0.1% and 1% of the constant power. These experiments use recordings of the total power measured on the photodiodes to infer the properties of the Optical Metrology System (OMS), such as reflectivities of the test masses and change of the photodiode efficiencies with time. In the first case the powers are back propagated from the different photodiodes to the same place on the optical bench to express the unknown quantities in the measurement with the complimentary photodiode measurements. They are combined in the way that the only unknown left is the test mass reflectivities. The second experiment compared two estimates of the force applied to the test masses due to the radiation pressure that appears because of the beam modulations. One estimate of the force is inferred from the measurements of the powers on the photodiodes and propagation of this measurement to the test masses. The other estimation of the force is done by calculating it from the change in the main scientific output of the instrument - differential displacement of the two test masses.

  14. Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

    2000-08-01

    The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

  15. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  16. Use of expert systems in the structural safety assessment of of pressurized nuclear components

    International Nuclear Information System (INIS)

    Jovanovic, A.; Sturm, D.

    1990-01-01

    The paper describes research currently performed at MPA Stuttgart on development of expert systems and application of artificial intelligence methods and techniques, for structural safety assessment of power plant pressurized components. The research is done as an extension of preceding and existing large research programs of MPA, in the domain of structural safety of components. In this preceding research a waste amount of practical engineering knowledge and experience has been accumulated: development in the direction of AI-based systems is a way to use this knowledge more efficiently in future research and in the nuclear power plant practice. Applications on which the current research is focussed are expert systems applied for the leak-before-break analysis for the structural safety evaluation in high temperature regimes

  17. Safety and reliability of pressure components with special emphasis on advanced methods of NDT. Vol. 2

    International Nuclear Information System (INIS)

    1986-01-01

    The 12 papers discuss topics of strength and safety in the field of materials technology and engineering. Conclusions for NPP component safety and materials are drawn. Measurements and studies relate to fracture mechanics methods (oscillation, burst, material strength, characteristics). The dynamic analysis of the behaviour of large test specimens, the influence of load velocity on crack resistance curve and the development of forged parts from austenitic steel for fast breeder reactors are presented. (DG) [de

  18. Present status of the disk pressure tests for hydrogen embrittlements

    International Nuclear Information System (INIS)

    Fidelle, J.P.

    1988-01-01

    The Disk Pressure Tests (DPT) have been developed considerably. Theoretically: a finite elements mechanical analysis shows the build up of a triaxial stress state already at the beginning of the test, which, with other reasons accounts for the very high sensitivity. Experimentally: for Internal Hydrogen Embrittlement (IHE) e.g. Co, Ti, U alloys, for environment embrittlement due to H 2 hydrogenated media such as water vapor, alcohol, machining fluids or liquid NH 3 . The range has been expanded considerably: up to 300 MPa and up to 1000 0 C. Very low strain rate - longer than a month - tests have been able to evidence HGE; of FCC alloys where H diffusivity is low for very oxidation -sensitive metals such as Nb and Ta, effects may appear only at somewhat high rates. The relationship between dynamic tests, static and low-cycle fatigue tests has been determined. In a number of instances, including SCC, other techniques and even fracture mechanics have been compared to the DPT and proved at best equivalent and several times, less sensitive than a well conducted DPT. At extreme they could not reproduce the field service phenomenon whereas the DPT did and could also be applied satisfactorily to low yield stress materials. The main rupture aspects have been analysed mechanically and organized in a rational and comprehensive chart based on 12,000 + tests over 15O + materials in different conditions. Comparison of HGE tests and service behaviour of a large variety of materials and industrial equipments has made possible to specify acceptance criteria for industrial service, which, provided the shape of the stress strain curves is not significantly affected, can be expanded to IHE, HE by other fluids than H 2 , 36 refs

  19. Certification test for safety of new fuel transportation package

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Sugawa, Osami; Suga, Masao.

    1993-01-01

    The objective of this certification test is to prove the safety of new fuel transportation package against a fire of actual size caused by traffic accidents. After the fire test, the fuel assemblies were covered with coal-tar like material vaporized from anti-shock material used in the container. Surface color of BWR-type fuel assembly was dark grey that is supposed to be the color of oxide of Zircaloy. As for PWR-type fuel assembly, the condition encountered during fire test caused no change to the outlook of the rod element. Both the BWR and PWR type fuel rod elements showed no deformation and were completely sound. Therefore it may be concluded that the container protected the mimic fuel assemblies against fire of 30 minutes duration and caused no damage. This report is the result of the above experiments and examinations, and we appreciate the cooperation of those who are concerned. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  1. Supplementary safety system 1/4 scale testing

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, R.L.; Paik, I.K.

    1993-09-01

    During the course of updating the K-Reactor Safety Analysis Report Chapter 15 in 1990, it was identified that the current Supplementary Safety System (SSS) may not be adequate in protecting the reactor during the process water pump coastdown initiated by a loss of AC power when the safety rods are assumed to fail. A SSS modification project was initiated to add an additional ink injection pathway near the pump suction. In addition, the Department of Energy raised a question on the thermal buoyancy effects on moderator flow pattern and ink dispersion in the moderator space. The development and documentation of a two-dimensional code called MODFLOW was undertaken to describe the problem. This report discusses the results of the moderator flow and ink (Gadolinium Poison Solution - GPS) dispersion tests designed to provide qualified data for validation and benchmarking of the MODFLOW computer code with the secondary objectives being the development of concentration profiles and video footage of simulated GPS dispersion under steady-state and transient flow conditions.

  2. Reliability assessment for safety critical systems by statistical random testing

    International Nuclear Information System (INIS)

    Mills, S.E.

    1995-11-01

    In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs

  3. Reliability assessment for safety critical systems by statistical random testing

    Energy Technology Data Exchange (ETDEWEB)

    Mills, S E [Carleton Univ., Ottawa, ON (Canada). Statistical Consulting Centre

    1995-11-01

    In this report we present an overview of reliability assessment for software and focus on some basic aspects of assessing reliability for safety critical systems by statistical random testing. We also discuss possible deviations from some essential assumptions on which the general methodology is based. These deviations appear quite likely in practical applications. We present and discuss possible remedies and adjustments and then undertake applying this methodology to a portion of the SDS1 software. We also indicate shortcomings of the methodology and possible avenues to address to follow to address these problems. (author). 128 refs., 11 tabs., 31 figs.

  4. Ambient Pressure Test Rig Developed for Testing Oil-Free Bearings in Alternate Gases and Variable Pressures

    Science.gov (United States)

    Bauman, Steven W.

    1990-01-01

    The Oil-Free Turbomachinery research team at the NASA Glenn Research Center is conducting research to develop turbomachinery systems that utilize high-speed, high temperature foil (air) bearings that do not require an oil lubrication system. Such systems combine the most advanced foil bearings from industry with NASA-developed hightemperature solid-lubricant technology. New applications are being pursued, such as Oil- Free turbochargers, auxiliary power units, and turbine propulsion systems for aircraft. An Oil-Free business jet engine, for example, would be simpler, lighter, more reliable, and less costly to purchase and maintain than current engines. Another application is NASA's Prometheus mission, where gas bearings will be required for the closed-cycle turbine based power-conversion system of a nuclear power generator for deep space. To support these applications, Glenn's Oil-Free Turbomachinery research team developed the Ambient Pressure Test Rig. Using this facility, researchers can load and heat a bearing and evaluate its performance with reduced air pressure to simulate high altitude conditions. For the nuclear application, the test chamber can be purged with gases such as helium to study foil gas bearing operation in working fluids other than air.

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

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  6. Vibration phenomena in large scale pressure suppression tests

    International Nuclear Information System (INIS)

    Aust, E.; Boettcher, G.; Kolb, M.; Sattler, P.; Vollbrandt, J.

    1982-01-01

    Structure und fluid vibration phenomena (acceleration, strain; pressure, level) were observed during blow-down experiments simulating a LOCA in the GKSS full scale multivent pressure suppression test facility. The paper describes first the source related excitations during the two regimes of condensation oscillation and of chugging, and deals then with the response vibrations of the facility's wetwell. Modal analyses of the wetwell were run using excitation by hammer and by shaker in order to separate phenomena that are particular to the GKSS facility from more general ones, i.e. phenomena specific to the fluid related parameters of blowdown and to the geometry of the vent pipes only. The lowest periodicities at about 12 and 16 Hz stem from the vent acoustics. A frequency of about 36 to 38 Hz prominent during chugging seems to result from the lowest local models of two of the wetwell's walls when coupled by the wetwell pool. Further peaks found during blowdown in the spectra of signals at higher frequencies correspond to global vibration modes of the wetwell. (orig.)

  7. JRR-3 cold neutron source facility H2-O2 explosion safety proof testing

    International Nuclear Information System (INIS)

    Hibi, T.; Fuse, H.; Takahashi, H.; Akutsu, C.; Kumai, T.; Kawabata, Y.

    1990-01-01

    A cold Neutron Source (CNS) will be installed in Japan Research Reactor-3 (JRR-3) in Japan Atomic Energy Research Institute (JAERI) during its remodeling project. This CNS holds liquid hydrogen at a temperature of about 20 K as a cold neutron source moderator in the heavy water area of the reactor to moderate thermal neutrons from the reactor to cold neutrons of about 5 meV energy. In the hydrogen circuit of the CNS safety measures are taken to prevent oxygen/hydrogen reaction (H 2 -O 2 explosion). It is also designed in such manner that, should an H 2 -O 2 explosion take place, the soundness of all the components can be maintained so as not to harm the reactor safety. A test hydrogen circuit identical to that of the CNS (real components designed by TECHNICATOME of France) was manufactured to conduct the H 2 -O 2 explosion test. In this test, the detonation that is the severest phenomenon of the oxygen/hydrogen reaction took place in the test hydrogen circuit to measure the exerted pressure on the components and their strain, deformation, leakage, cracking, etc. Based on the results of this measurement, the structural strength of the test hydrogen circuit was analyzed. The results of this test show that the hydrogen circuit components have sufficient structural strength to withstand an oxygen/hydrogen reaction

  8. LOFT integral test system final safety analysis report

    International Nuclear Information System (INIS)

    1974-03-01

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

  9. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  10. An experimental study on the thermal-hydraulic phenomena in the Hybrid Safety Injection Tank using a separate effect test facility

    International Nuclear Information System (INIS)

    Ryu, Sung Uk; Ryu, Hyobong; Park, Hyun-Sik; Yi, Sung-Jae

    2016-01-01

    Highlights: • The experimental study on the pressure balancing between the Hybrid SIT and PZR. • The effects of different variables affecting the pressure balancing are investigated. • A sensitivity analysis on the pressure variations of the Hybrid SIT. - Abstract: This paper reports an experimental research for investigating thermal hydraulic phenomena of Hybrid Safety Injection Tank (Hybrid SIT) using a separate effect test facility in Korea Atomic Energy Research Institute (KAERI). The Hybrid SIT is a passive safety injection system that enables the safety injection water to be injected into the reactor pressure vessel throughout all operating pressures by connecting the top of the SIT and the pressurizer (PZR). The separate effect test (SET) facility of Hybrid SIT, which is designed based on the APR+ power plant, comprises a PZR, Hybrid SIT, pressure balancing line (PBL), injection line (IL), nitrogen gas line, and refueling water tank (RWT). Furthermore, the pressure loss range of the SET facility was analyzed and compared with that of the reference nuclear power plant. In this research, a condition for balancing the pressure between the Hybrid SIT and PZR is examined and the effects of different variables affecting the pressure balancing, which are flow rate, injection velocity of steam and initial water level, are also investigated. The condition for balancing the pressure between the Hybrid SIT and PZR was derived theoretically from a pressure network for the Hybrid SIT, pressurizer, and reactor pressure vessel. Additionally, a sensitivity analysis as a theoretical approach was conducted on the pressure variations in relation to the rate of steam condensation inside the Hybrid SIT. The results showed that pressure of the Hybrid SIT was predominantly determined by the rate of steam condensation. The results showed that if the rate of condensation increased or decreased by 10%, the Hybrid SIT pressure at the pressure balancing point decreased or

  11. 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.

  12. 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.

  13. 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

  14. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  15. Safety Significance of the Halden IFA-650 LOCA Test Results

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Petit, Marc; Hozer, Zoltan; Kelppe, Seppo; Khvostov, Grigori; Hafidi, Biya; Therache, Benjamin; Heins, Lothar; Valach, Mojmir; Voglewede, John; Wiesenack, Wolfgang

    2010-01-01

    The safety criteria for loss-of-coolant accidents were defined to ensure that the core would remain coolable. Since the time of the first LOCA experiments, which were largely conducted with fresh fuel, changes in fuel design, the introduction of new cladding materials and in particular the move to high burnup have generated a need to re-examine these criteria and to verify their continued validity. As part of international efforts to this end, the OECD Halden Reactor Project program implemented a LOCA test series. Based on recommendations of a group of experts from the US NRC, EPRI, EDF, IRSN, FRAMATOME-ANP and GNF, the primary objective of the experiments were defined as 1. Measure the extent of fuel (fragment) relocation into the ballooned region and evaluate its possible effect on cladding temperature and oxidation. 2. Investigate the extent (if any) of 'secondary transient hydriding' on the inner side of the cladding above and below the burst region. The fourth test of the series, IFA-650.4 conducted in April 2006, caused particular attention in the international nuclear community. The fuel used in the experiment had a high burnup, 92 MWd/kgU, and a low pre-test hydrogen content of about 50 ppm. The test aimed at and achieved a peak cladding temperature of 850 deg. C. The rod burst occurred at 790 deg. C. The burst caused a marked temperature increase at the lower end and a decrease at the upper end of the system, indicating that fuel relocation had occurred. Subsequent gamma scanning showed that approximately 19 cm of the fuel stack were missing from the upper part of the rod and that fuel had fallen to the bottom of the capsule. PIE at the IFE-Kjeller hot cells corroborated this evidence of substantial fuel relocation. The fact that fuel dispersal could occur upon ballooning and burst, i.e. at cladding temperatures as low as 800 deg. C and thus far lower than the temperature entailed by the current 1200 deg. C / 17% ECR limit, caused concern. The

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

    International Nuclear Information System (INIS)

    Heorman, K.R.

    1992-01-01

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

  17. Test tools of physics radiography children as a support for safety radiation and safety patients

    International Nuclear Information System (INIS)

    Siti Masrochah; Yeti Kartikasari; Ardi Soesilo Wibowo

    2013-01-01

    Radiographic examination of the thorax children aged 1-3 years have a high sufficiently failure. This failure is caused by the movement and difficulty positioning the patient, resulting in the risk of repeat radiographs to patient safety particularly unnecessary radiation risks. It is therefore necessary to develop research on children design fixation devices. This research aims to create a design tool fixation on radiographs children to support radiation safety and patient safety. This research is a descriptive exploratory approach to tool design. The independent variables were the design tools, variable tool function test results, and radiographic variables controlled thorax. The procedure is done by designing data collection tools, further trials with 20 samples. Processing and analysis of data is done by calculating the performance assessment tool scores with range 1-3. The results showed that the design tool of fixation in the form of standard radiographic cassette equipped with chairs and some form of seat belt fixation. The procedure uses a tool fixation is routine radiographic follow thorax child in an upright position. Function test results aids fixation is to have an average score of 2.66, which means good. While the test results for each component, the majority of respondents stated that the reliability of the device is quite good with a score of 2.45 (60 %), convenience tool with a score of 2.60 (70 %), quality of the radiographs did not incontinence of the thorax radiograph with a score 2.55 (85 %), the child protection (security) with a score of 2.70 (70 %), good design aesthetic design with a score of 2.80 (80 %), addition of radiation from the others on the use of these tools do not need with a score of 2.80 (80 %), and there is no additional radiation due to repetitions with a score of 2.85 (90 %). (author)

  18. Applications of high and ultra high pressure homogenization for food safety

    Directory of Open Access Journals (Sweden)

    Francesca Patrignani

    2016-08-01

    Full Text Available Traditionally, the shelf-life and safety of foods have been achieved by thermal processing. Low temperature long time (LTLT and high temperature short time (HTST treatments are the most commonly used hurdles for the pasteurization of fluid foods and raw materials. However, the thermal treatments can reduce the product quality and freshness. Consequently, some non-thermal pasteurization process have been proposed during the last decades, including high hydrostatic pressure (HHP, pulsed electric field (PEF, ultrasound (US and high pressure homogenization (HPH. This last technique has been demonstrated to have a great potential to provide fresh-like products with prolonged shelf-life. Moreover, the recent developments in high-pressure-homogenization technology and the design of new homogenization valves able to withstand pressures up to 350-400 MPa have opened new opportunities to homogenization processing in the food industries and, consequently, permitted the development of new products differentiated from traditional ones by sensory and structural characteristics or functional properties. For this, this review deals with the principal mechanisms of action of high pressure homogenization against microorganisms of food concern in relation to the adopted homogenizer and process parameters. In addition, the effects of homogenization on foodborne pathogenic species inactivation in relation to the food matrix and food chemico-physical and process variables will be reviewed. Also the combined use of this alternative technology with other non-thermal technologies will be considered

  19. The safety of food products requires X-ray testing

    International Nuclear Information System (INIS)

    Lardiere, C.

    2017-01-01

    Food safety through standards and regulations imposes food products to be tested for the presence of alien elements. So far metal detectors have been used to detect metal parts, now they have been progressively replaced with X-ray scanners that allow the detection of a lot more contaminants. The improvement of algorithms for image processing combined with the availability of ever more powerful PC have led to the routine use of X-ray testing on industrial processes. Technological progress has made X-ray testing more efficient: while previously a power of 500 W was necessary to cross a 10 cm thickness of water, now only 100 W is necessary. The main advantage of X-ray testing is to be able to test food even packed in metal containers and to detect if the container is deformed. Another advantages is to be able to detect a lot of elements like pieces of glass, small stones or bits of bones. The minimal size to be detected is 0.5 mm for stainless steels and 2 mm for glass or bones. Usually metal detectors are set at the end of the production line just before packaging but in some cases they can be included in the line in order to protect the next machine that intervenes to process the food. (A.C.)

  20. Safety of a rapid diagnostic protocol with accelerated stress testing.

    Science.gov (United States)

    Soremekun, Olan A; Hamedani, Azita; Shofer, Frances S; O'Conor, Katie J; Svenson, James; Hollander, Judd E

    2014-02-01

    Most patients at low to intermediate risk for an acute coronary syndrome (ACS) receive a 12- to 24-hour "rule out." Recently, trials have found that a coronary computed tomographic angiography-based strategy is more efficient. If stress testing were performed within the same time frame as coronary computed tomographic angiography, the 2 strategies would be more similar. We tested the hypothesis that stress testing can safely be performed within several hours of presentation. We performed a retrospective cohort study of patients presenting to a university hospital from January 1, 2009, to December 31, 2011, with potential ACS. Patients placed in a clinical pathway that performed stress testing after 2 negative troponin values 2 hours apart were included. We excluded patients with ST-elevation myocardial infarction or with an elevated initial troponin. The main outcome was safety of immediate stress testing defined as the absence of death or acute myocardial infarction (defined as elevated troponin within 24 hours after the test). A total of 856 patients who presented with potential ACS were enrolled in the clinical pathway and included in this study. Patients had a median age of 55.0 (interquartile range, 48-62) years. Chest pain was the chief concern in 86%, and pain was present on arrival in 73% of the patients. There were no complications observed during the stress test. There were 0 deaths (95% confidence interval, 0%-0.46%) and 4 acute myocardial infarctions within 24 hours (0.5%; 95% confidence interval, 0.14%-1.27%). The peak troponins were small (0.06, 0.07, 0.07, and 0.19 ng/mL). Patients who present to the ED with potential ACS can safely undergo a rapid diagnostic protocol with stress testing. © 2013.

  1. On the transient pressure build-up in the full pressure safety shell of watercooled nuclear reactors after a loss of coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1979-08-01

    The thermo-and fluid-dynamic processes in a multichamber full pressure safety containment during a loss of coolant accident have been investigated. Comparison of the calculations carried out with the computer programs, in which ZOCO VI was used as being representative of similar programs, with the experimental results pointed out discrepancies in the determination of time dependent pressure, pressure difference and temperature curves. This led to the development of a new theoretical model and a program COFLOW which pays particular attention to the fluid dynamic processes in the initial phase of a loss of coolant accident. It can also be used to determine the maximum containment pressure towards the end of a loss of coolant accident. Comparison of the COFLOW results with experiments has shown that COFLOW provides a model and a procedure by which the physical processes in a multichamber full pressure safety containment can be simulated satisfactorily

  2. 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...

  3. Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

    International Nuclear Information System (INIS)

    Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G.; Souto, Joao P.R.S.; Carvalho Junior, Ideir T.

    2013-01-01

    Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

  4. Residual heat removal pump and low pressure safety injection pump retrofit program

    International Nuclear Information System (INIS)

    Dudiak, J.G.; McKenna, J.M.

    1992-01-01

    Residual Heat Removal (RHR) and low pressure safety injection (LPSI) pumps installed in pressurized water-to-reactor power plants are used to provide low-head safety injection in the event of loss of coolant in the reactor coolant system. Because these pumps are subjected to rather severe temperature and pressure transients, the majority of pumps installed in the RHR service are vertical pumps with a single stage impeller. Typically the pump impeller is mounted on an extended motor shaft (close-coupled configuration) and a mechanical seal is employed at the pump end of the shaft. Traditionally RHR and LPSI pumps have been a significant maintenance item for many utilities. Periodic mechanical seal of motor bearing replacement often is considered routine maintenance. The closed-coupled pump design requires disassembly of the casing cover from the lower pump casing while performing these routine maintenance tasks. This paper introduces a design modification developed to convert the close-coupled RHR and LPSI pumps to a coupled configuration

  5. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

  6. Test description and preliminary pitot-pressure surveys for Langley Test Technique Demonstrator at Mach 6

    Science.gov (United States)

    Everhart, Joel L.; Ashby, George C., Jr.; Monta, William J.

    1992-01-01

    A propulsion/airframe integration experiment conducted in the NASA Langley 20-Inch Mach 6 Tunnel using a 16.8-in.-long version of the Langley Test Technique Demonstrator configuration with simulated scramjet propulsion is described. Schlieren and vapor screen visualization of the nozzle flow field is presented and correlated with pitot-pressure flow-field surveys. The data were obtained at nominal free-stream conditions of Re = 2.8 x 10 exp 6 and a nominal engine total pressure of 100 psia. It is concluded that pitot-pressure surveys coupled to schlieren and vapor-screen photographs, and oil flows have revealed flow features including vortices, free shear layers, and shock waves occurring in the model flow field.

  7. Standard Test Method for Saltwater Pressure Immersion and Temperature Testing of Photovoltaic Modules for Marine Environments

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method provides a procedure for determining the ability of photovoltaic modules to withstand repeated immersion or splash exposure by seawater as might be encountered when installed in a marine environment, such as a floating aid-to-navigation. A combined environmental cycling exposure with modules repeatedly submerged in simulated saltwater at varying temperatures and under repetitive pressurization provides an accelerated basis for evaluation of aging effects of a marine environment on module materials and construction. 1.2 This test method defines photovoltaic module test specimens and requirements for positioning modules for test, references suitable methods for determining changes in electrical performance and characteristics, and specifies parameters which must be recorded and reported. 1.3 This test method does not establish pass or fail levels. The determination of acceptable or unacceptable results is beyond the scope of this test method. 1.4 The values stated in SI units are to be ...

  8. A modified isometric test to evaluate blood pressure control with ...

    African Journals Online (AJOL)

    Blood pressure at rest is not predictive of roundthe- clock values. Blood pressure should therefore be measured during effort to evaluate hypertension and its response to treatment. The effect of sustained-release verapamil (240 mg taken once a day) on blood pressure at rest and during isometric effort was therefore ...

  9. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author)

  10. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author).

  11. Development of Onsite Transportation Safety Documents for Nevada Test Site

    International Nuclear Information System (INIS)

    Frank Hand; Willard Thomas; Frank Sciacca; Manny Negrete; Susan Kelley

    2008-01-01

    Department of Energy (DOE) Orders require each DOE site to develop onsite transportation safety documents (OTSDs). The Nevada Test Site approach divided all onsite transfers into two groups with each group covered by a standalone OTSD identified as Non-Nuclear and Nuclear. The Non-Nuclear transfers involve all radioactive hazardous material in less than Hazard Category (HC)-3 quantities and all chemically hazardous materials. The Nuclear transfers involve all radioactive material equal to or greater than HC-3 quantities and radioactive material mated with high explosives regardless of quantity. Both OTSDs comply with DOE O 460.1B requirements. The Nuclear OTSD also complies with DOE O 461.1A requirements and includes a DOE-STD-3009 approach to hazard analysis (HA) and accident analysis as needed. All Nuclear OTSD proposed transfers were determined to be non-equivalent and a methodology was developed to determine if 'equivalent safety' to a fully compliant Department of Transportation (DOT) transfer was achieved. For each HA scenario, three hypothetical transfers were evaluated: a DOT-compliant, uncontrolled, and controlled transfer. Equivalent safety is demonstrated when the risk level for each controlled transfer is equal to or less than the corresponding DOT-compliant transfer risk level. In this comparison the typical DOE-STD-3009 risk matrix was modified to reflect transportation requirements. Design basis conditions (DBCs) were developed for each non-equivalent transfer. Initial DBCs were based solely upon the amount of material present. Route-, transfer-, and site-specific conditions were evaluated and the initial DBCs revised as needed. Final DBCs were evaluated for each transfer's packaging and its contents

  12. 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

  13. 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.

  14. Inconsistent food safety pressures complicate environmental conservation for California produce growers

    Directory of Open Access Journals (Sweden)

    Patrick Baur

    2016-08-01

    Full Text Available Controlling human pathogens on fresh vegetables, fruits and nuts is imperative for California growers. A range of rules and guidelines have been developed since 2006, when a widespread outbreak of E. coli O157:H7 was linked to bagged spinach grown in California. Growers face pressure from industry and government sources to adopt specific control measures on their farms, resulting in a complex, shifting set of demands, some of which conflict with environmental stewardship. We surveyed 588 California produce growers about on-farm practices related to food safety and conservation. Nearly all respondents considered both food safety and environmental protection to be important responsibilities for their farms. Responses indicate that clearing vegetation to create buffers around cropped fields, removing vegetation from ditches and ponds, and using poison bait and wildlife fences are commonly used practices intended to reduce wildlife movements onto farm fields. The survey also revealed that on-farm practices vary substantially even among farms with similar characteristics. This variability suggests inconsistencies in food safety requirements, auditors' interpretations or growers' perception of the demands of their buyers. Although site-specific considerations are important and practices should be tailored to local conditions, our findings suggest growers, natural resources and food safety would benefit from clearer, more consistent requirements.

  15. Safety conditions of using structural steels under high temperature and pressures in hydrogen containing environment

    International Nuclear Information System (INIS)

    Asviyan, M.B.

    1984-01-01

    The method for establishing full-strength conditions was suggested on the base of results of creep-rupture test of tube samples under hydrogen pressure and according to permissible stresses in neutral medium. Applicability of the method was considered taking St3 and 12KhM steels as examples. It was shown that the use of suggested dependences and special efficiency factors enables to forecast endurance limit for the given steel grade and assigned partial hydrogen pressure without labour-intensive test conducting

  16. Environmental testing of an experimental digital safety channel

    International Nuclear Information System (INIS)

    Korsah, K.; Tanaka, T.J.; Wilson, T.L. Jr.; Wood, R.T.

    1996-09-01

    This document presents the results of environmental stress tests performed on an experimental digital safety channel (EDSC) assembled at the Oak Ridge National Laboratory (ORNL) as part of the NRC-sponsored Qualification of Advanced Instrumentation and Controls (W) System program. The objective of this study is to investigate failure modes and vulnerabilities of microprocessor-based technologies when subjected to environmental stressors. The study contributes to the technical basis for environmental qualification of safety-related digital I ampersand C systems. The EDSC employs technologies and digital subsystems representative of those proposed for use in advanced light-water reactors (ALWRs) or for retrofits in existing plants. Subsystems include computers, electrical and optical serial communication links, fiber-optic network links, analog-to-digital and digital-to-analog converters, and multiplexers. The EDSC was subjected to selected stressors that are a potential risk to digital equipment in a mild environment. The selected stressors were electromagnetic and radio-frequency interference (EMYRFI), temperature, humidity, and smoke exposure. The stressors were applied over ranges that were considerably higher than what the channel is likely to experience in a normal nuclear power plant environment. Ranges of stress were selected at a sufficiently high level to induce errors so that failure modes that are characteristic of the technologies employed could be identified

  17. Safety procedures in operation of inspection and maintenance of pressure reduction and metering stations

    International Nuclear Information System (INIS)

    Villas Boas, Ademar Jose; Biesemeyer, Marco Aurelio R.

    2000-01-01

    Each local Natural Gas Distribution Company in Brazil has its own working procedures for operations of inspection and maintenance on equipment and accessories connected to the gas network. Some of these Companies developed a better elaborated and documented way of working routines, while others only work based on their operators experience. The objective of this work is to create a standard procedure for operations of inspection and maintenance of Pressure Reducing Stations and Metering Stations, mainly the ones concerned to safety aspects. This work has no intention of exhausting all aspects related to this subject but to become the first step to standardize these types of operations among Natural Gas Distribution Companies. (author)

  18. Safety analysis of Atucha 1 reactor pressure vessel for a typical transient

    International Nuclear Information System (INIS)

    Chomik, E.; Jinchuk, D.

    1994-01-01

    As a consequence of disturbances on the CNA I external electric grid some incidents were produced in a 6 minutes lapse, causing a sudden cooling of the primary system, while pressure was maintained nearly constant. On the basis of this event, a safety analysis based on the LInear Elastic Fracture Mechanics was carried out. This paper presents an alternative method for the calculation of transients; the Finite Element Method, particularly, the OCA-II FEM code. By using this method it was possible to demonstrate, for this event, a safe operating condition for the end of life of the RPV, with regard to brittle fracture risk. 6 refs, 11 figs, 1 tab

  19. Research of explosives in an environment of high pressure and temperature using a new test stand

    Directory of Open Access Journals (Sweden)

    Jan Drzewiecki

    2015-01-01

    Full Text Available In this article the test stand for determining the blast abilities of explosives in high pressure and temperature conditions as well as the initial results of the research are presented. Explosives are used in rock burst and methane prevention to destroy precisely defined fragments of the rock mass where energy and methane are accumulated. Using this preventive method for fracturing the structure of the rocks which accumulate the energy or coal of the methane seam very often does not bring the anticipated results. It is because of the short range of destructive action of the post-blast gases around the blast hole. Evaluation of the blast dynamics of explosives in a test chamber, i.e. in the pressure and temperature conditions comparable to those found “in situ”, will enable evaluation of their real usefulness in commonly used mining hazard preventive methods. At the same time, it will enable the development of new designs of the explosive charges used for precisely determined mining hazards. In order to test the explosives for their use in difficult environmental conditions and to determine the characteristics of their explosion, a test chamber has been built. It is equipped with a system of sensors and a high-frequency recording system of pressure and temperature during a controlled explosion of an explosive charge. The results of the research will enable the development of new technologies for rock burst and methane prevention which will significantly increase workplace health and safety level. This paper presented results constitute the initial phase of research started in the middle of 2014.

  20. Safety implications associated with in-plant pressurized gas storage and distribution systems

    International Nuclear Information System (INIS)

    Guymon, R.H.

    1986-01-01

    Storage and handling of compressed gases at nuclear power plants were studied to identify any potential safety hazards. Gases investigated were air, acetylene, carbon dioxide, chlorine, Halon, hydrogen, nitrogen, oxygen, propane, and sulfur hexafluoride. Physical properties of gases were reviewed, as were applicable industrial codes and standards. Incidents involving pressurized gases in general industry and in the nuclear industry were studied. In this report general hazards, such as missiles from ruptures, rocketing of cylinders, fires, explosions, asphyxiation, and toxicity, are discussed. Even though some serious injuries and deaths have occurred over the years in industries handling and using pressurized gases, the industrial codes, standards, practices, and procedures are very comprehensive. The most important step one can take to ensure the safe handling of gases is to enforce these well-known and established methods

  1. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

  2. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    1990-04-01

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  3. Testing of Full Scale Flight Qualified Kevlar Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Greene, Nathanael; Saulsberry, Regor; Yoder, Tommy; Forsyth, Brad; Thesken, John; Phoenix, Leigh

    2007-01-01

    Many decades ago NASA identified a need for low-mass pressure vessels for carrying various fluids aboard rockets, spacecraft, and satellites. A pressure vessel design known as the composite overwrapped pressure vessel (COPV) was identified to provide a weight savings over traditional single-material pressure vessels typically made of metal and this technology has been in use for space flight applications since the 1970's. A typical vessel design consisted of a thin liner material, typically a metal, overwrapped with a continuous fiber yarn impregnated with epoxy. Most designs were such that the overwrapped fiber would carry a majority of load at normal operating pressures. The weight advantage for a COPV versus a traditional singlematerial pressure vessel contributed to widespread use of COPVs by NASA, the military, and industry. This technology is currently used for personal breathing supply storage, fuel storage for auto and mass transport vehicles and for various space flight and aircraft applications. The NASA Engineering and Safety Center (NESC) was recently asked to review the operation of Kevlar 2 and carbon COPVs to ensure they are safely operated on NASA space flight vehicles. A request was made to evaluate the life remaining on the Kevlar COPVs used on the Space Shuttle for helium and nitrogen storage. This paper provides a review of Kevlar COPV testing relevant to the NESC assessment. Also discussed are some key findings, observations, and recommendations that may be applicable to the COPV user community. Questions raised during the investigations have revealed the need for testing to better understand the stress rupture life and age life of COPVs. The focus of this paper is to describe burst testing of Kevlar COPVs that has been completed as a part of an the effort to evaluate the effects of ageing and shelf life on full scale COPVs. The test articles evaluated in this discussion had a diameter of 22 inches for S/N 014 and 40 inches for S/N 011. The

  4. Safety demonstration test (SR-1/S1C-1) plan of HTTR (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Shigeaki; Sakaba, Nariaki; Takada, Eiji; Tachibana, Yukio; Saito, Kenji; Furusawa, Takayuki; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2003-03-01

    Safety demonstration tests in the HTTR (High Temperature Engineering Test Reactor) will be carried out in order to verify inherent safety features of the HTGR (High Temperature Gas-cooled Reactor). The first phase of the safety demonstration tests includes the reactivity insertion test by the control rod withdrawal and the coolant flow reduction test by the circulator trip. In the second phase, accident simulation tests will be conducted. By comparison of their experimental and analytical results, the prediction capability of the safety evaluation codes such as the core and the plant dynamics codes will be improved and verified, which will contribute to establish the safety design and the safety evaluation technologies of the HTGRs. The results obtained through its safety demonstration tests will be also utilised for the establishment of the safety design guideline, the safety evaluation guideline, etc. This paper describes the test program of the overall safety demonstration tests and the test method, the test conditions and the results of the pre-test analysis of the reactivity insertion test and the partial gas circulator trip test planned in March 2003. (author)

  5. Safety test No. S-6, launch pad abort sequential test Phase II: solid propellant fire

    International Nuclear Information System (INIS)

    Snow, E.C.

    1975-08-01

    In preparation for the Lincoln Laboratory's LES 8/9 space mission, a series of tests was performed to evaluate the nuclear safety capability of the Multi-Hundred Watt (MHW) Radioisotope Thermoelectric Generator (RTG) to be used to supply power for the satellite. One such safety test is Test No. S-6, Launch Pad Abort Sequential Test. The objective of this test was to subject the RTG and its components to the sequential environments characteristic of a catastrophic launch pad accident to evaluate their capability to contain the 238 PuO 2 fuel. This sequence of environments was to have consisted of the blast overpressure and fragments, followed by the fireball, low velocity impact on the launch pad, and solid propellant fire. The blast overpressure and fragments were subsequently eliminated from this sequence. The procedures and results of Phase II of Test S-6, Solid Propellant Fire are presented. In this phase of the test, a simulant Fuel Sphere Assembly (FSA) and a mockup of a damaged Heat Source Assembly (HSA) were subjected to single proximity solid propellant fires of approximately 10-min duration. Steel was introduced into both tests to simulate the effects of launch pad debris and the solid rocket motor (SRM) casing that might be present in the fire zone. (TFD)

  6. 42 CFR 84.163 - Man test for gases and vapors; Type C supplied-air respirators, demand and pressure-demand...

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Man test for gases and vapors; Type C supplied-air respirators, demand and pressure-demand classes; test requirements. 84.163 Section 84.163 Public Health PUBLIC HEALTH SERVICE, DEPARTMENT OF HEALTH AND HUMAN SERVICES OCCUPATIONAL SAFETY AND HEALTH RESEARCH AND RELATED ACTIVITIES APPROVAL OF...

  7. The Fast Flux Test Facility built on safety

    International Nuclear Information System (INIS)

    1989-01-01

    No other high-tech industry has grown as fast as the nuclear industry. The information available to the general public has not kept pace with the rapid growth of nuclear data---its growth has outpaced its media image and the safety of nuclear facilities has become a highly debated issue. This book is an attempt to bridge the gap between the high-tech information of the nuclear industry and its understanding by the general public. It explains the three levels of defense at the Fast Flux Test Facility (FFTF) and why these levels provide an acceptable margin to protect the general public and on-site personnel, while achieving FFTF's mission to provide research and development for the US Department of Energy

  8. Very high temperature measurements: Applications to nuclear reactor safety tests

    International Nuclear Information System (INIS)

    Parga, Clemente-Jose

    2013-01-01

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: - The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  9. Applications of High and Ultra High Pressure Homogenization for Food Safety.

    Science.gov (United States)

    Patrignani, Francesca; Lanciotti, Rosalba

    2016-01-01

    Traditionally, the shelf-life and safety of foods have been achieved by thermal processing. Low temperature long time and high temperature short time treatments are the most commonly used hurdles for the pasteurization of fluid foods and raw materials. However, the thermal treatments can reduce the product quality and freshness. Consequently, some non-thermal pasteurization process have been proposed during the last decades, including high hydrostatic pressure, pulsed electric field, ultrasound (US), and high pressure homogenization (HPH). This last technique has been demonstrated to have a great potential to provide "fresh-like" products with prolonged shelf-life. Moreover, the recent developments in high-pressure-homogenization technology and the design of new homogenization valves able to withstand pressures up to 350-400 MPa have opened new opportunities to homogenization processing in the food industries and, consequently, permitted the development of new products differentiated from traditional ones by sensory and structural characteristics or functional properties. For this, this review deals with the principal mechanisms of action of HPH against microorganisms of food concern in relation to the adopted homogenizer and process parameters. In addition, the effects of homogenization on foodborne pathogenic species inactivation in relation to the food matrix and food chemico-physical and process variables will be reviewed. Also the combined use of this alternative technology with other non-thermal technologies will be considered.

  10. Safety analysis of the 700-horsepower combustion test facility

    Energy Technology Data Exchange (ETDEWEB)

    Berkey, B.D.

    1981-05-01

    The objective of the program reported herein was to provide a Safety Analysis of the 700 h.p. Combustion Test Facility located in Building 93 at the Pittsburgh Energy Technology Center. Extensive safety related measures have been incorporated into the design, construction, and operation of the Combustion Test Facility. These include: nitrogen addition to the coal storage bin, slurry hopper, roller mill and pulverizer baghouse, use of low oxygen content combustion gas for coal conveying, an oxygen analyzer for the combustion gas, insulation on hot surfaces, proper classification of electrical equipment, process monitoring instrumentation and a planned remote television monitoring system. Analysis of the system considering these factors has resulted in the determination of overall probabilities of occurrence of hazards as shown in Table I. Implementation of the recommendations in this report will reduce these probabilities as indicated. The identified hazards include coal dust ignition by hot ductwork and equipment, loss of inerting within the coal conveying system leading to a coal dust fire, and ignition of hydrocarbon vapors or spilled oil, or slurry. The possibility of self-heating of coal was investigated. Implementation of the recommendations in this report will reduce the ignition probability to no more than 1 x 10/sup -6/ per event. In addition to fire and explosion hazards, there are potential exposures to materials which have been identified as hazardous to personal health, such as carbon monoxide, coal dust, hydrocarbon vapors, and oxygen deficient atmosphere, but past monitoring experience has not revealed any problem areas. The major environmental hazard is an oil spill. The facility has a comprehensive spill control plan.

  11. Leak detection of SF6 gas pressure vessel safety devices at BARC-TIFR Pelletron accelerator

    International Nuclear Information System (INIS)

    Ninawe, N.G.; Sharma, S.C.; Nair, J.P.; Sparrow, H.; Bolar, P.C.; Gudekar, P.V.; Mahapatra, S.; Vishwakarma, R.S.; Ramjilal; Matkar, U.V.; Gore, J.A.; Gupta, A.K.

    2015-01-01

    Pelletron accelerator is in operation since last more than 26 years. To achieve desired voltage gradient SF6 gas of about 20 tons is used to have 75-80 psig pressure in main accelerator tank. During accelerator tank maintenance gas is transferred to four storage tanks, kept in open space in the vicinity of sea. Recently refurbishment and retrofitting of four storage tanks was carried out which includes the installation of new drift space, rupture disc assembly, relief valves and manual valves along with civil and painting work. All components to be installed were tested for high pressure. Helium gas sniffer technique was used to check micro leaks for new joints for all components before installing for storage tanks. Subsequently, the tanks were tested up to 90 psig SF6 gradually in succession. No pressure drop was observed in storage tanks. This work was carried out as per recommendation of the then particle accelerator committee (PASC). (author)

  12. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs

  13. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs.

  14. Safety and reliability of pressure components with special emphasis on advanced methods of NDT. Vol. 1

    International Nuclear Information System (INIS)

    1986-01-01

    24 papers discuss various methods for nondestructive testing of materials, e.g. eddy current measurement, EMAG analyser, tomography, ultrasound, holographic interferometry, and optical sound field camera. Special consideration is given to mathematical programmes and tests allowing to determine fracture-mechanical parameters and to assess cracks in various components, system parts and individual specimens both in pressurized systems and NPP systems. Studies focus on weld seams and adjacent areas. (DG) [de

  15. 76 FR 34145 - Safety Zone, Barrier Testing Operations, Chicago Sanitary and Ship Canal, Romeoville, IL

    Science.gov (United States)

    2011-06-13

    ...-AA00 Safety Zone, Barrier Testing Operations, Chicago Sanitary and Ship Canal, Romeoville, IL AGENCY.... Construction on Barrier IIB has been completed. Operational and safety testing was conducted in February 2011... dispersal barrier IIA and IIB. This safety zone will be enforced daily from 7 a.m. to 11 a.m. and from 1 p.m...

  16. Managing blood pressure control in Asian patients: safety and efficacy of losartan.

    Science.gov (United States)

    Cheung, Tommy Tsang; Cheung, Bernard Man Yung

    2014-01-01

    Hypertension is common in Asian populations and is a major cause of cardiovascular diseases. The prevalence of hypertension is increasing in many Asian countries. The overall prevalence of hypertension in India and the People's Republic of China has been estimated to be 20.6% in men and 22.6% in women. However, the rates of detection, treatment, and control of hypertension remain low in Asia. This reflects a low level of literacy and education, as well as a low level of access to medical care. To overcome these obstacles, strategies targeted at education, promotion, and optimization of medical care, are crucial to achieve target blood pressure control. Angiotensin receptor blockers are one of the first-line treatments for essential hypertension because they confer better cardiovascular outcomes. Losartan has been widely evaluated for the management of hypertension. Although some studies suggested that the blood pressure-lowering effect of losartan is perhaps lower than for other angiotensin receptor blockers, losartan has been demonstrated to be beneficial in terms of renal protection in patients with diabetes, heart failure resulting from either systolic or diastolic dysfunction, and diuretic-induced hyperuricemia. However, most of these data were obtained from Caucasian populations. The efficacy and safety of losartan in Asian populations may be different because of genetic and ethnic variations. Therefore, the efficacy and safety of losartan in Asian patients with hypertension warrant further study.

  17. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    Energy Technology Data Exchange (ETDEWEB)

    Burdin, J.R.

    1995-10-25

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  18. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    Burdin, J.R.

    1995-01-01

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  19. Test Structures for Rapid Prototyping of Gas and Pressure Sensors

    Science.gov (United States)

    Buehler, M.; Cheng, L. J.; Martin, D.

    1996-01-01

    A multi-project ceramic substrate was used in developing a gas sensor and pressure sensor. The ceramic substrate cantained 36 chips with six variants including sensors, process control monitors, and an interconnect ship. Tha gas sensor is being developed as an air quality monitor and the pressure gauge as a barometer.

  20. Proof pressure tests of the PCPVs at Hinkley Point B and Hunterston B

    International Nuclear Information System (INIS)

    Eadie, D.McD.; Bell, D.J.

    1976-01-01

    The two PCPVs at Hinkley Point B were pressure tested in August 1973 and April 1974. The first vessel at Hunterston B was tested in December, 1973, and the second early in 1975. The vessels were pressurised up to 709 psig and, at various stages of pressurisation, readings were taken of external deflections, internal corner deflections and concrete strains. Surveys were taken of external concrete cracks and of crack gauges embedded in the concrete near the re-entrant corners. The external vessel deflections were measured optically using telescopes, targets and invar tapes. In some cases a recent design of manometric equipment was used to monitor the vertical deflections of the top slab during pressurisation and at proof pressure when access to the vessel was not possible for safety reasons. Internal concrete strains were measured using vibrating wire strain gauges. A brief description is given of the various measuring devices used. Deflection readings were also taken of some penetration primary closures. Summaries of the various recorded readings are given and compared with the design analyses. (author)

  1. Finite test sets development method for test execution of safety critical software

    International Nuclear Information System (INIS)

    Shin, Sung Min; Kim, Hee Eun; Kang, Hyun Gook; Lee, Sung Jiun

    2014-01-01

    The V and V method has been utilized for this safety critical software, while SRGM has difficulties because of lack of failure occurrence data on developing phase. For the safety critical software, however, failure data cannot be gathered after installation in real plant when we consider the severe consequence. Therefore, to complement the V and V method, the test-based method need to be developed. Some studies on test-based reliability quantification method for safety critical software have been conducted in nuclear field. These studies provide useful guidance on generating test sets. An important concept of the guidance is that the test sets represent 'trajectories' (a series of successive values for the input variables of a program that occur during the operation of the software over time) in the space of inputs to the software.. Actually, the inputs to the software depends on the state of plant at that time, and these inputs form a new internal state of the software by changing values of some variables. In other words, internal state of the software at specific timing depends on the history of past inputs. Here the internal state of the software which can be changed by past inputs is named as Context of Software (CoS). In a certain CoS, a software failure occurs when a fault is triggered by some inputs. To cover the failure occurrence mechanism of a software, preceding researches insist that the inputs should be a trajectory form. However, in this approach, there are two critical problems. One is the length of the trajectory input. Input trajectory should long enough to cover failure mechanism, but the enough length is not clear. What is worse, to cover some accident scenario, one set of input should represent dozen hours of successive values. The other problem is number of tests needed. To satisfy a target reliability with reasonable confidence level, very large number of test sets are required. Development of this number of test sets is a herculean

  2. Ultrasonic data acquisition installation for basis and in-service testing of nuclear pressure vessels

    International Nuclear Information System (INIS)

    Gutmann, G.; Engl, G.

    1976-01-01

    The safety of nuclear installations requires continuous safety inspections during construction and operation. Essential parts of this safety inspection are the basis and in-line inspections. For this purpose installation systems are used which allow an optimal statement to be made regarding the conditions of tested components

  3. Thermal-Hydraulic Tests for Reactor Core Safety

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

    2005-04-01

    The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

  4. Influence of different safety shoes on gait and plantar pressure: a standardized examination of workers in the automotive industry.

    Science.gov (United States)

    Ochsmann, Elke; Noll, Ulrike; Ellegast, Rolf; Hermanns, Ingo; Kraus, Thomas

    2016-09-30

    Working conditions, such as walking and standing on hard surfaces, can increase the development of musculoskeletal complaints. At the interface between flooring and musculoskeletal system, safety shoes may play an important role in the well-being of employees. The aim of this study was to evaluate the effects of different safety shoes on gait and plantar pressure distributions on industrial flooring. Twenty automotive workers were individually fitted out with three different pairs of safety shoes ( "normal" shoes, cushioned shoes, and midfoot bearing shoes). They walked at a given speed of 1.5 m/s. The CUELA measuring system and shoe insoles were used for gait analysis and plantar pressure measurements, respectively. Statistical analysis was conducted by ANOVA analysis for repeated measures. Walking with cushioned safety shoes or a midfoot bearing safety shoe led to a significant decrease of the average trunk inclination (pshoes as well as midfoot bearing shoes (pshoes. As expected, plantar pressure distributions varied significantly between cushioned or midfoot bearing shoes and shoes without ergonomic components. The overall function of safety shoes is the avoidance of injury in case of an industrial accident, but in addition, safety shoes could be a long-term preventive instrument for maintaining health of the employees' musculoskeletal system, as they are able to affect gait parameters. Further research needs to focus on safety shoes in working situations.

  5. Comparison between monitored and modeled pore water pressure and safety factor in a slope susceptible to shallow landslides

    Science.gov (United States)

    Bordoni, Massimiliano; Meisina, Claudia; Zizioli, Davide; Valentino, Roberto; Bittelli, Marco; Chersich, Silvia

    2014-05-01

    Shallow landslides can be defined as slope movements affecting superficial deposits of small thicknesses which are usually triggered due to extreme rainfall events, also very concentrated in time. Shallow landslides are hazardous phenomena: in particular, if they happen close to urbanized areas they could cause significant damages to cultivations, structures, infrastructures and, sometimes, human losses. The triggering mechanism of rainfall-induced shallow landslides is strictly linked with the hydrological and mechanical responses of usually unsaturated soils to rainfall events. For this reason, it is fundamental knowing the intrinsic hydro-mechanical properties of the soils in order to assess both susceptibility and hazard of shallow landslide and to develop early-warning systems at large scale. The hydrological data collected by a 20 months monitoring on a slope susceptible to shallow landslides in an area of the North -Eastern Oltrepo Pavese (Northern Apennines, Italy) were used to identify the hydrological behaviors of the investigated soils towards rainfall events. Field conditions under different rainfall trends have also been modeled by using both hydrological and physically-based stability models for the evaluation of the slope safety factor . The main objectives of this research are: (a) to compare the field measured pore water pressures at different depths with results of hydrological models, in order to evaluate the efficiency of the tested models and to determine how precipitations affect pore pressure development; (b) to compare the time trends of the safety factor that have been obtained by applying different stability models; (c) to evaluate, through a sensitivity analysis, the effects of soil hydrological properties on modeling pore water pressure and safety factor. The test site slope where field measurements were acquired is representative of other sites in Northern Apennines affected by shallow landslides and is characterized by medium

  6. Design of experiments and equipment to test the ballooning characteristics of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Forrest, C.F.; Stern, F.; Hart, R.G.

    1992-01-01

    Experiments have been planned and an apparatus has been designed to enable creep testing of end-of-life pressure tube specimens in a LOCA environment. Effects that could be studied include: annealing of irradiation damage during transient heating; effects of hydride blisters on pressure tube ballooning strains; and, effects of uniformly-distributed hydrogen content on pressure tube ballooning strains. The proposed experimental program will consist of separate effects creep tests on pressure tube sections under transient heating conditions

  7. Environmental safety analysis tests on the Light Weight Radioisotope Heater Unit (LWRHU)

    International Nuclear Information System (INIS)

    Tate, R.E.; Land, C.C.

    1985-05-01

    A series of safety tests has been performed on the Light Weight Radioisotope Heater Unit (LWRHU), a 238 PuO 2 -fueled device designed to provide thermal energy at selected locations in a spacecraft. The tests simulate the thermal and mechanical environments postulated for spacecraft accidents on the launch pad and on reentry abort. The tests demonstrate almost complete containment of the fuel, or fuel simulant (depleted UO 2 ), in (1) an overpressure environment of 12.76 MPa (1850 psi), (2) on impact by an 18-g aluminum fuel-tank fragment at velocities greater than 750 m/s (2460 ft/s) but less than 900 m/s (2950 ft/s), (3) during a 10.5-min burn of a 0.9 x 0.9 x 0.9 m (3 x 3 x 3 ft) block of solid rocket motor propellant, (4) after impact at 49 m/s (161 ft/s) in four different orientations on a hard surface, and (5) during immersion in seawater for 1.75 years at both sea level pressure and at a pressure equivalent to 6000 m (19,700 ft) of ocean depth

  8. Study for Relation of Pressure and Aging Degradation during LOCA Test

    International Nuclear Information System (INIS)

    Kim, Jong Seog

    2013-01-01

    As result of this test, it was found that low pressure effect in aging was not significant compared with that of temperature. If temperature profile in LOCA test can satisfy the plant LOCA profile, no further analysis of pressure profile for aging degradation is necessary. For environmental qualification of electric equipment in containment building of nuclear power plant, LOCA test should be applied. During the LOCA test, temperature and pressure of LOCA chamber shall be controlled to meet a requirement of plant specific LOCA profile. It is general to keep LOCA test temperature and pressure above the plant specific LOCA profile. If the test temperature is lower than required profile in some time zone while it is higher in other time zone, calculation of total cumulated test temperature is required to compare with that of plant profile. Arrhenius equation can be applied for calculation of total temperature accumulation. If there is a deviation of pressure between test profile and plant specific profile, can we still use the same rule of temperature? Since the Arrhenius equation can't be applied to pressure, analysis of pressure effect to aging degradation is not easy. Study for relation of pressure and aging degradation during LOCA condition is described herein. To Study an aging degradation effect of pressure during LOCA test, comparison of IR during high LOCA pressure and low LOCA pressure were implemented. We expected low IR in high pressure because it contained a high concentration of oxygen which induces high aging degradation. Contrary to our expectation, IR of low pressure was lower than that of high pressure. It is assumed that high vibration of temperature profile to maintain the low pressure at high temperature induced supply of high enthalpy steam into LOCA chamber

  9. Estimation of the Blood Pressure Response With Exercise Stress Testing.

    Science.gov (United States)

    Fitzgerald, Benjamin T; Ballard, Emma L; Scalia, Gregory M

    2018-04-20

    The blood pressure response to exercise has been described as a significant increase in systolic BP (sBP) with a smaller change in diastolic BP (dBP). This has been documented in small numbers, in healthy young men or in ethnic populations. This study examines these changes in low to intermediate risk of myocardial ischaemia in men and women over a wide age range. Consecutive patients having stress echocardiography were analysed. Ischaemic tests were excluded. Manual BP was estimated before and during standard Bruce protocol treadmill testing. Patient age, sex, body mass index (BMI), and resting and peak exercise BP were recorded. 3200 patients (mean age 58±12years) were included with 1123 (35%) females, and 2077 males, age range 18 to 93 years. Systolic BP increased from 125±17mmHg to 176±23mmHg. The change in sBP (ΔsBP) was 51mmHg (95% CI 51,52). The ΔdBP was 1mmHg (95% CI 1, 1), from 77 to 78mmHg, p<0.001). The upper limit of normal peak exercise sBP (determined by the 90th percentile) was 210mmHg in males and 200mmHg in females. The upper limit of normal ΔsBP was 80mmHg in males and 70mmHg in females. The lower limit of normal ΔsBP was 30mmHg in males and 20mmHg in females. In this large cohort, sBP increased significantly with exercise. Males had on average higher values than females. Similar changes were seen with the ΔsBP. The upper limit of normal for peak exercise sBP and ΔsBP are reported by age and gender. Copyright © 2018 Australian and New Zealand Society of Cardiac and Thoracic Surgeons (ANZSCTS) and the Cardiac Society of Australia and New Zealand (CSANZ). All rights reserved.

  10. Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

    Science.gov (United States)

    Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

    2018-01-01

    Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

  11. Formal testing and utilization of streaming media to improve flight crew safety knowledge.

    Science.gov (United States)

    Bellazzini, Marc A; Rankin, Peter M; Quisling, Jason; Gangnon, Ronald; Kohrs, Mike

    2008-01-01

    Increased concerns over the safety of air medical transport have prompted development of novel ways to increase safety. The objective of our study was to determine if an Internet streaming media safety video increased crew safety knowledge. 23 out of 40 crew members took an online safety pre-test, watched a safety video specific to our program and completed immediate and long-term post-testing 6 months later. Mean pre-test, post-test and 6 month follow up test scores were 84.9%, 92.3% and 88.4% respectively. There was a statistically significant difference in all scores (p Streaming media proved to be an accessible and effective supplement to safety training in our study.

  12. Uniform relativistic universe models with pressure. Part 2. Observational tests

    International Nuclear Information System (INIS)

    Krempec, J.; Krygier, B.

    1977-01-01

    The magnitude-redshift and angular diameter-redshift relations are discussed for the uniform (homogeneous and isotropic) relativistic Universe models with pressure. The inclusion of pressure into the energy-momentum tensor has given larger values of the deceleration parameter q. An increase of the deceleration parameter has led to the brightening of objects as well as to a little larger angular diameters. (author)

  13. Adjustable guide for a testing system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Seifert, W.

    1980-01-01

    The device consisting of a guide rail and a manipulator is introduced into the gap between pressure vessel wall and biological shield by means of suspending wire drums and manipulator drums. For adjustment of the device an elbow telescope is used. The guide rail is fixed to the pressure vessel wall by means of electromagnets. The movements of the manipulator with respect to the guide rail are performed with the aid of a motor. (DG) [de

  14. High pressure cells for magnetic measurements - destruction and functional tests

    Czech Academy of Sciences Publication Activity Database

    Kamarád, Jiří; Machátová, Zuzana; Arnold, Zdeněk

    2004-01-01

    Roč. 75, č. 11 (2004), s. 5022-5025 ISSN 0034-6748 R&D Projects: GA ČR GA202/02/0739; GA AV ČR IAA1010315 Institutional research plan: CEZ:AV0Z1010914 Keywords : pressure cells * pressure transmitting media * CuBe * MP35N Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.226, year: 2004

  15. Preliminary development of an advanced modular pressure relief cushion: Testing and user evaluation.

    Science.gov (United States)

    Freeto, Tyler; Mitchell, Steven J; Bogie, Kath M

    2018-02-01

    Effective pressure relief cushions are identified as a core assistive technology need by the World Health Organization Global Cooperation on Assistive Technology. High quality affordable wheelchair cushions could provide effective pressure relief for many individuals with limited access to advanced assistive technology. Value driven engineering (VdE) principles were employed to develop a prototype modular cushion. Low cost dynamically responsive gel balls were arranged in a close packed array and seated in bilayer foam for containment and support. Two modular cushions, one with high compliance balls and one with moderate compliance balls were compared with High Profile and Low Profile Roho ® and Jay ® Medical 2 cushions. ISO 16480-2 biomechanical standardized tests were applied to assess cushion performance. A preliminary materials cost analysis was carried out. A prototype modular cushion was evaluated by 12 participants who reported satisfaction using a questionnaire based on the Quebec User Evaluation of Satisfaction with Assistive Technology (QUEST 2.0) instrument. Overall the modular cushions performed better than, or on par with, the most widely prescribed commercially available cushions under ISO 16480-2 testing. Users rated the modular cushion highly for overall appearance, size and dimensions, comfort, safety, stability, ease of adjustment and general ease of use. Cost-analysis indicated that every modular cushion component a could be replaced several times and still maintain cost-efficacy over the complete cushion lifecycle. A VdE modular cushion has the potential provide effective pressure relief for many users at a low lifetime cost. Copyright © 2017. Published by Elsevier Ltd.

  16. PA171 Containers on a Wood Pallet with Metal Top Adapter, Air Pressure Tests During MIL-STD-1660 Tests

    National Research Council Canada - National Science Library

    2004-01-01

    ... (PM-MAS) to conduct Air Pressure Tests during MIL-STD-1660, "Design Criteria for Ammunition Unit Loads" testing on the PA171 containers on a wood pallet with metal top adapter as manufactured by Alliant Tech...

  17. Prediction of safety critical software operational reliability from test reliability using testing environment factors

    International Nuclear Information System (INIS)

    Jung, Hoan Sung; Seong, Poong Hyun

    1999-01-01

    It has been a critical issue to predict the safety critical software reliability in nuclear engineering area. For many years, many researches have focused on the quantification of software reliability and there have been many models developed to quantify software reliability. Most software reliability models estimate the reliability with the failure data collected during the test assuming that the test environments well represent the operation profile. User's interest is however on the operational reliability rather than on the test reliability. The experiences show that the operational reliability is higher than the test reliability. With the assumption that the difference in reliability results from the change of environment, from testing to operation, testing environment factors comprising the aging factor and the coverage factor are developed in this paper and used to predict the ultimate operational reliability with the failure data in testing phase. It is by incorporating test environments applied beyond the operational profile into testing environment factors. The application results show that the proposed method can estimate the operational reliability accurately. (Author). 14 refs., 1 tab., 1 fig

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

    International Nuclear Information System (INIS)

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

    2006-06-01

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

  19. Fast Flux Test Facility final safety analysis report. Amendment 72

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D. A.

    1992-08-01

    This document provides the Final Safety Analysis Report (FSAR) Amendment 72 for incorporation into the Fast Flux Test Facility (FFTF) FSAR set. This amendment change incorporates Engineering Change Notices issued subsequent to Amendment 71 and approved for incorporation before June 24, 1992. These include changes in: Chapter 2, Site Characteristics; Chapter 3, Design Criteria Structures, Equipment, and Systems; Chapter 5B, Reactor Coolant System; Chapter 7, Instrumentation and Control Systems; Chapter 8, Electrical Systems - The description of the Class 1E, 125 Vdc systems is updated for the higher capacity of the newly installed, replacement batteries; Chapter 9, Auxiliary Systems - The description of the inert cell NASA systems is corrected to list the correct number of spare sample points; Chapter 11, Reactor Refueling System; Chapter 12, Radiation Protection and Waste Management; Chapter 13, Conduct of Operations; Chapter 16, Quality Assurance; Chapter 17, Technical Specifications; Chapter 19, FFTF Fire Specifications for Fire Detection, Alarm, and Protection Systems; Chapter 20, FFTF Criticality Specifications; and Appendix B, Primary Piping Integrity Evaluation.

  20. Safety testing of GM-rice expressing PHA-E lectin using a new animal test design

    DEFF Research Database (Denmark)

    Poulsen, Morten; Schrøder, Malene; Wilcks, Andrea

    2007-01-01

    The 90-day animal study is the core study for the safety assessment of genetically modified foods in the SAFOTEST project. The model compound tested in the 90-day study was a rice variety expressing the kidney bean Phaseolus vulgaris lectin agglutinin E-form (PHA-E lectin). Female Wistar rats were...... safety testing of genetically modified foods....

  1. On-line testing of response time and calibration of temperature and pressure sensors in nuclear power plants

    International Nuclear Information System (INIS)

    Hashemian, H.M.

    1995-01-01

    Periodic calibrations and response time measurements are necessary for temperature and pressure sensors in the safety systems of nuclear power plants. Conventional measurement methods require the test to be performed at the sensor location or involve removing the sensor from the process and performing the tests in a laboratory or on the bench. The conventional methods are time consuming and have the potential of causing wear and tear on the equipment, can expose the test personnel to radiation and other harsh environments, and increase the length of the plant outage. Also, the conventional methods do not account for the installation effects which may have an influence on sensor performance. On-line testing methods alleviate these problems by providing remote sensor response time and calibration capabilities. For temperature sensors such as Resistance Temperature Detectors (RTDs) and thermocouples, an on-line test method called the Loop Current Step Response (LCSR) technique has been developed, and for pressure transmitters, an on-line method called noise analysis which was available for reactor diagnostics was validated for response time testing applications. Both the LCSR and noise analysis tests are performed periodically in U.S. nuclear power plants to meet the plant technical specification requirements for response time testing of safety-related sensors. Automated testing of the calibration of both temperature and pressure sensors can be accomplished through an on-line monitoring system installed in the plant. The system monitors the DC output of the sensors over the fuel cycle to determine if any calibration drift has occurred. Changes in calibration can be detected using signal averaging and intercomparison methods and analytical redundancy techniques. (author)

  2. 33 CFR 183.580 - Static pressure test for fuel tanks.

    Science.gov (United States)

    2010-07-01

    ... pressure test for fuel tanks. A fuel tank is tested by performing the following procedures in the following... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Static pressure test for fuel tanks. 183.580 Section 183.580 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND...

  3. Pressure sensor for use in the Loss-of-Fluid-Test (LOFT) reactor

    International Nuclear Information System (INIS)

    Billeter, T.R.

    1979-07-01

    Tests at temperatures up to 800 0 F and pressures up to 2500 psig were conducted at Hanford Engineering Development Laboratory (HEDL) to qualify an instrument for measurement of fuel-rod pressure in the Loss-of-Fluid-Test (LOFT) reactor. Operational characteristics of the selected pressure transducers are summarized for a series of static, quasi-static, and transient tests conducted for a period of about 700 hours

  4. Wind pressure testing of tornado safe room components made from wood

    Science.gov (United States)

    Robert Falk; Deepak Shrestha

    2016-01-01

    To evaluate the ability of a wood tornado safe room to resist wind pressures produced by a tornado, two safe room com-ponents were tested for wind pressure strength. A tornado safe room ceiling panel and door were static-pressure-tested according to ASTM E 330 using a vacuum test system. Re-sults indicate that the panels had load capacities from 2.4 to 3.5 times that...

  5. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    Science.gov (United States)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  6. DHS small-scale safety and thermal testing of improvised explosives-comparison of testing performance

    International Nuclear Information System (INIS)

    Reynolds, J G; Hsu, P C; Sandstrom, M M; Brown, G W; Warner, K F; Phillips, J J; Shelley, T J; Reyes, J A

    2014-01-01

    One of the first steps in establishing safe handling procedures for explosives is small-scale safety and thermal (SSST) testing. To better understand the response of improvised materials or homemade explosives (HMEs) to SSST testing, 16 HME materials were compared to three standard military explosives in a proficiency-type round robin study among five laboratories-two DoD and three DOE-sponsored by DHS. The testing matrix has been designed to address problems encountered with improvised materials-powder mixtures, liquid suspensions, partially wetted solids, immiscible liquids, and reactive materials. More than 30 issues have been identified that indicate standard test methods may require modification when applied to HMEs to derive accurate sensitivity assessments needed for developing safe handling and storage practices. This paper presents a generalized comparison of the results among the testing participants, comparison of friction results from BAM (German Bundesanstalt für Materi-alprüfung) and ABL (Allegany Ballistics Laboratory) designed testing equipment, and an overview of the statistical results from the RDX (1,3,5-Trinitroperhydro-1,3,5-triazine) standard tested throughout the proficiency test.

  7. 78 FR 54510 - New Entrant Safety Assurance Program Operational Test

    Science.gov (United States)

    2013-09-04

    ... safety management controls; (2) consider their effects on small businesses; and (3) consider establishing alternate locations where such reviews may be conducted for the convenience of small businesses. In response... safety review within 18 months of starting interstate operations. [49 U.S.C. 31144(g)]. In issuing these...

  8. Evaluation of safety test needs for the gas cooled breeder reactors

    International Nuclear Information System (INIS)

    Emon, D.E.; Buttemer, D.R.; Sevy, R.H.

    1976-01-01

    This paper deals with the process used in determining the safety test needs for the Gas Cooled Fast Breeder Reactor (GCFR), reports existing tentative conclusions, and indicates the direction that the process is taking at this time. The process is based upon two ideas: (1) that the safety information needs will be identified through risk analysis directly dependent on the various design features of the GCFR and (2) that the safety program will be determined by a safety review committee. The paper limits itself to presenting thoughts on the safety test needs directly associated with the GCFR core during severe beyond design basis accident situations involving the loss of coolable core geometry. Representative event sequence diagrams are reported for the three generic classes of accidents considered. The following categories of information are identified: safety information needs, safety tests required to fulfill these information needs, and the facilities required to perform the tests

  9. Structural safety of HDR reactor building during large scale vibration tests

    International Nuclear Information System (INIS)

    Stangenberg, F.; Zinn, R.

    1985-01-01

    In the second phase of the HDR investigations, a high shaker excitation of the building is planned using a large shaker which will be located on the operating floor and will be brought up to speed in a balanced condition and then unbalanced and decoupled from the drive system. With decreasing speed the shaker comes in resonance with the building frequencies and its energy is transferred to the building. In this paper the structural safety of the reactor building during the projected shaker tests is analysed. Dynamic response calculations with coupling between building and shaker by simultaneously integrating the equilibrium equations of both building and shaker are presented. The resulting building stresses, soil pressures etc. are compared with allowable values. (orig.)

  10. Safety aspects of using gadolinium as burnable poison in pressurized water reactors

    International Nuclear Information System (INIS)

    Vandenberg, C.; Bonet, H.; Charlier, A.; Motte, F.

    1979-01-01

    Within the framework of an experimental program on the behavior of gadolinium in light water reactors (LWRs), the BR3 power plant, a small 11-MW(electric) pressurized water reactor, was operated successfully with a core containing 5% Gd 2 O 3 -UO 2 rods. The core reached an average burnup increase of 22,000 MWd/tM, corresponding to 500 effective full-power days in a single cycle. These results were used to extrapolate the consequences on safety of extending such a control policy to large LWRs. In this context, the following factors were investigated: impact on the design, reactivity control and core behavior operated with lower and more constant boric acid concentration, environmental impact, fuel handling, etc

  11. Adolescents, gangs, and perceptions of safety, parental engagement, and peer pressure.

    Science.gov (United States)

    Kelly, Sarah E; Anderson, Debra G

    2012-10-01

    Adolescents are exposed to various forms of gang violence, and such exposure has led them to feel unsafe in their neighborhood and have differing interactions with their parents and peers. This qualitative study explored adolescents', parents', and community center employees' perceptions of adolescents' interaction with their neighborhood, family, and peers. Three themes emerged from the data: Most adolescents reported that the community center provided a safe environment for them; parental engagement influenced adolescents' experiences with gangs; and adolescents were subjected to peer pressure in order to belong. Exposure to gang violence can leave an impression on adolescents and affect their mental health, but neighborhood safety and relationships with parents and peers can influence adolescents' exposure to gang violence. Recommendations regarding the use of health care professionals at community centers are proposed. Copyright 2012, SLACK Incorporated.

  12. Application of a support vector machine algorithm to the safety precaution technique of medium-low pressure gas regulators

    Science.gov (United States)

    Hao, Xuejun; An, Xaioran; Wu, Bo; He, Shaoping

    2018-02-01

    In the gas pipeline system, safe operation of a gas regulator determines the stability of the fuel gas supply, and the medium-low pressure gas regulator of the safety precaution system is not perfect at the present stage in the Beijing Gas Group; therefore, safety precaution technique optimization has important social and economic significance. In this paper, according to the running status of the medium-low pressure gas regulator in the SCADA system, a new method for gas regulator safety precaution based on the support vector machine (SVM) is presented. This method takes the gas regulator outlet pressure data as input variables of the SVM model, the fault categories and degree as output variables, which will effectively enhance the precaution accuracy as well as save significant manpower and material resources.

  13. Status of researches in the field of safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel

    2017-01-01

    This collective publication proposes a synthesis of the status of researches performed in the field of safety of pressurized water reactors. They may discuss past, current and projected research works, involved actors, or lessons learned from these works. The authors propose a presentation of some research tools privileged by the IRSN for these researches: the CABRI and PHEBUS reactors, the GALAXIE experimental platform, and some other installations. Then they address researches related to loss-of-coolant accidents (two-phase thermohydraulics, fuel rod behaviour), to reactivity accidents, to accidents related to dewatering of irradiated fuel storage pools, to fires, to extreme aggressions of natural origin (earthquake, extreme flooding), to core fusion accidents (core heating and fusion within the vessel, vessel failure and apron erosion by corium, containment enclosure dynamic loading, release of radioactive products), to the behaviour of nuclear plant important metallic or civil works components and notably to their ageing, to organisational and human factors or more generally to social and human sciences (design of control rooms, safety organisation and management in EDF nuclear plants), and to other issues and research perspectives

  14. Use of a risk assessment method to improve the safety of negative pressure wound therapy.

    Science.gov (United States)

    Lelong, Anne-Sophie; Martelli, Nicolas; Bonan, Brigitte; Prognon, Patrice; Pineau, Judith

    2014-06-01

    To conduct a risk analysis of the negative pressure wound therapy (NPWT) care process and to improve the safety of NPWT, a working group of nurses, hospital pharmacists, physicians and hospital managers performed a risk analysis for the process of NPWT care. The failure modes, effects and criticality analysis (FMECA) method was used for this analysis. Failure modes and their consequences were defined and classified as a function of their criticality to identify priority actions for improvement. By contrast to classical FMECA, the criticality index (CI) of each consequence was calculated by multiplying occurrence, severity and detection scores. We identified 13 failure modes, leading to 20 different consequences. The CI of consequences was initially 712, falling to 357 after corrective measures were implemented. The major improvements proposed included the establishment of 6-monthly training cycles for nurses, physicians and surgeons and the introduction of computerised prescription for NPWT. The FMECA method also made it possible to prioritise actions as a function of the criticality ranking of consequences and was easily understood and used by the working group. This study is, to our knowledge, the first to use the FMECA method to improve the safety of NPWT. © 2012 The Authors. International Wound Journal © 2012 Medicalhelplines.com Inc and John Wiley & Sons Ltd.

  15. Application of probabilistic fracture mechanics to reactor pressure vessel safety assessment

    International Nuclear Information System (INIS)

    Venturini, V.; Pitner, P.

    1995-06-01

    Among all the components of a PWR (Pressurized Water Reactor) nuclear power plant, the reactor vessel is of major importance for safety. The integrity of this structure must be guaranteed in all circumstances, even in the case of the most severe accidents, and its mechanical state can be decisive for the lifetime of the plant. The brittle rupture would be the most important of all potential hazards if the irradiation effects were not consistent with predictions. The interest of having a reliable and precise method of evaluating the available safety margins and the integrity of this component led Electricite de France (EDF) to carry out a probabilistic fracture mechanics analysis. The probabilistic model developed by integration of the uncertainties in the usual fracture mechanics equations is presented. A special focus is made on the problem of coupling thermo-mechanical finite element calculations and reliability analysis. The use of a finite element code can be associated with prohibitive computation times when it is invoked numerous times during simulations sequences or complex iterative procedures. The response surface method is used. It provides an approximation of the response from a reduced number of original data. The global approach is illustrated on an example corresponding to a specific accidental transient. A validation of the obtained results is also carried out through the comparison with an equivalent model without coupling. (author)

  16. Technology, safety, and costs of decommissioning a reference pressurized water reactor power station

    International Nuclear Information System (INIS)

    Smith, R.I.; Konzek, G.J.; Kennedy, W.E. Jr.

    1978-05-01

    Safety and cost information was developed for the conceptual decommissioning of a large [1175 MW(e)] pressurized water reactor (PWR) power station. Two approaches to decommissioning, Immediate Dismantlement and Safe Storage with Deferred Dismantlement, were studied to obtain comparisons between costs, occupational radiation doses, potential radiation dose to the public, and other safety impacts. Immediate Dismantlement was estimated to require about six years to complete, including two years of planning and preparation prior to final reactor shutdown, at a cost of $42 million, and accumulated occupational radiation dose, excluding transport operations, of about 1200 man-rem. Preparations for Safe Storage were estimated to require about three years to complete, including 1 1 / 2 years for planning and preparation prior to final reactor shutdown, at a cost of $13 million and an accumulated occupational radiation dose of about 420 man-rem. The cost of continuing care during the Safe Storage period was estimated to be about $80 thousand annually. Accumulated occupational radiation dose during the Safe Storage period was estimated to range from about 10 man-rem for the first 10 years to about 14 man-rem after 30 years or more. The cost of decommissioning by Safe Storage with Deferred Dismantlement was estimated to be slightly higher than Immediate Dismantlement. Cost reductions resulting from reduced volumes of radioactive material for disposal, due to the decay of the radioactive containments during the deferment period, are offset by the accumulated costs of surveillance and maintenance during the Safe Storage period

  17. Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

    International Nuclear Information System (INIS)

    Podkopaev, V.; Popov, V.; Zaritsky, N.

    1997-01-01

    This emergency situation had occurred at the ZNPP unit 1 while its being under ''hot shutdown'' in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ''Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs

  18. Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

    Energy Technology Data Exchange (ETDEWEB)

    Podkopaev, V; Popov, V; Zaritsky, N [State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Kiev (Ukraine)

    1997-09-01

    This emergency situation had occurred at the ZNPP unit 1 while its being under ``hot shutdown`` in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ``Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  20. Test results of Run-1 and Run-2 in steam generator safety test facility (SWAT-3)

    International Nuclear Information System (INIS)

    Kurihara, A.; Yatabe, Toshio; Tanabe, Hiromi; Hiroi, Hiroshi

    2003-07-01

    Large leak sodium-water reaction tests were carried out using SWAT-1 rig and SWAT-3 facility in Power Reactor and Nuclear Fuel Development Corporation (PNC) O-arai Engineering Center to obtain the data on the design of the prototype LMFBR Monju steam generator against a large leak accident. This report provides the results of SWAT-3 Runs 1 and 2. In Runs 1 and 2, the heat transfer tube bundle of the evaporator, fabricated by TOSHIBA/IHI, were used, and the pressure relief line was located at the top of evaporator. The water injection rates in the evaporator were 6.7 kg/s and 14.2 (initial)-9.7 kg/s in Runs 1 and 2 respectively, which corresponded to 3.3 tubes and 7.1 (initial)-4.8 tubes failure in actual size system according to iso-velocity modeling. Approximately two hundreds of measurement points were provided to collect data such as pressure, temperature, strain, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on in each run. Initial spike pressures were 1.13 MPa and 2.62 MPa nearest to injection point in Runs 1 and 2 respectively, and the maximum quasi-steady pressures in evaporator were 0.49 MPa and 0.67 MPa in Runs 1 and 2. No secondary tube failure was observed. The rupture disc of evaporator (RD601) burst at 1.1s in Run-1 and at 0.7s in Run-2 after water injected, and the pressure relief system was well-functioned though a few items for improvement were found. (author)

  1. Reducing uncertainty in geostatistical description with well testing pressure data

    Energy Technology Data Exchange (ETDEWEB)

    Reynolds, A.C.; He, Nanqun [Univ. of Tulsa, OK (United States); Oliver, D.S. [Chevron Petroleum Technology Company, La Habra, CA (United States)

    1997-08-01

    Geostatistics has proven to be an effective tool for generating realizations of reservoir properties conditioned to static data, e.g., core and log data and geologic knowledge. Due to the lack of closely spaced data in the lateral directions, there will be significant variability in reservoir descriptions generated by geostatistical simulation, i.e., significant uncertainty in the reservoir descriptions. In past work, we have presented procedures based on inverse problem theory for generating reservoir descriptions (rock property fields) conditioned to pressure data and geostatistical information represented as prior means for log-permeability and porosity and variograms. Although we have shown that the incorporation of pressure data reduces the uncertainty below the level contained in the geostatistical model based only on static information (the prior model), our previous results assumed did not explicitly account for uncertainties in the prior means and the parameters defining the variogram model. In this work, we investigate how pressure data can help detect errors in the prior means. If errors in the prior means are large and are not taken into account, realizations conditioned to pressure data represent incorrect samples of the a posteriori probability density function for the rock property fields, whereas, if the uncertainty in the prior mean is incorporated properly into the model, one obtains realistic realizations of the rock property fields.

  2. Design Improvement of Double Pressure Vessel in the In-pile Test Section

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Heo, Sung-Ho; Joung, Chang-Young; Kim, Ka-Hye [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To carry out an irradiation test of nuclear fuels, a nuclear fuel test rig should be fabricated and installed in the in-pile test section (IPS), which is installed in the reactor hall. While carrying out an irradiation test, sealing out coolant which passes through the test rig is one of the most important issues. In particular, although the double pressure vessel is assembled with the IPS head by two o-rings and six bolts, 15.5 MPa of highly pressurized coolant leaks through the gap between the vessel and IPS head. Because the temperature of the coolant in the test loop is 300 .deg. C , and the pool of HANARO is 40 .deg. C, the double pressure vessel is necessary to insulate them. Therefore, a new design to prevent the leakage of coolant needs to be developed. In this study, EB welding technique is considered to assemble the double pressure vessel and the IPS head, and their mechanical design is modified to enable the welding process. In this study, an improved design for sealing out the coolant at the pressure boundary between the double pressure vessel and the IPS head has been developed. An EB weld is applied to seal out the pressure boundary, and its sealing performance is verified by NDE, a cross section test, and a hydraulic pressure test. From the verification test results, the improved design can be used in fabricating the IPS for a nuclear fuel irradiation test.

  3. Safety report content and development for test loop facility on MARIA reactor

    International Nuclear Information System (INIS)

    Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

    1982-01-01

    A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

  4. Improvement of auditing technology of safety analysis through thermal-hydraulic separate effect tests

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Moon, Young Min; Lee, Dong Won; Lee, Sang Ik; Kim, Eung Soo; Yeom, Keum Soo [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-03-15

    The objective of the present research is to perform the separate effect tests and to assess the RELAP5/MOD3.2 code for the analysis of thermal-hydraulic behavior in the reactor coolant system and the improvement of the auditing technology of safety analysis. Three Separate Effect Tests (SETs) are the reflux condensation in the U-tube, the direct contact condensation in the hot-leg and the mixture level buildup in the pressurizer. The experimental data and the empirical correlations are obtained through SETs. On the ases of the three SET works, models in RELAP5 are modified and improved, which are compared with the data. The Korea Standard Nuclear Power Plant (KSNP) are assessed using the modified RELAP5. In the reflux condensation test, the data of heat transfer coefficients and flooding are obtained and the condensation models are modified using the non-iterative model, as results, modified code better predicts the data. In the direct contact condensation test, the data of heat transfer coefficients are obtained for the cocurrent and countercurrent flow between the mixture gas and the water in condition of horizontal stratified flow. Several condensation and friction models are modified, which well predict the present data. In the mixture level test, the data for the mixture level and the onset of water draining into the surge line are obtained. The standard RELAP5 over-predicts the mixture level and the void fraction in the pressurizer. Simple modification of model related to the pool void fraction is suggested. The KSNP is assessed using the standard and the modified RELAP5 resulting from the experimental and code works for the SETs. In case of the pressurizer manway opening with available secondary side of the steam generators, the modified code predicts that the collapsed level in the pressurizer is little accumulated. The presence and location of the opening and the secondary condition of the steam generators have an effect on the coolant inventory. The

  5. Influence of different safety shoes on gait and plantar pressure: a standardized examination of workers in the automotive industry

    OpenAIRE

    Ochsmann, Elke; Noll, Ulrike; Ellegast, Rolf; Hermanns, Ingo; Kraus, Thomas

    2016-01-01

    Objective: Working conditions, such as walking and standing on hard surfaces, can increase the development of musculoskeletal complaints. At the interface between flooring and musculoskeletal system, safety shoes may play an important role in the well-being of employees. The aim of this study was to evaluate the effects of different safety shoes on gait and plantar pressure distributions on industrial flooring. Methods: Twenty automotive workers were individually fitted out with three differe...

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. The predictive value of the sacral base pressure test in detecting specific types of sacroiliac dysfunction

    Science.gov (United States)

    Mitchell, Travis D.; Urli, Kristina E.; Breitenbach, Jacques; Yelverton, Chris

    2007-01-01

    Abstract Objective This study aimed to evaluate the validity of the sacral base pressure test in diagnosing sacroiliac joint dysfunction. It also determined the predictive powers of the test in determining which type of sacroiliac joint dysfunction was present. Methods This was a double-blind experimental study with 62 participants. The results from the sacral base pressure test were compared against a cluster of previously validated tests of sacroiliac joint dysfunction to determine its validity and predictive powers. The external rotation of the feet, occurring during the sacral base pressure test, was measured using a digital inclinometer. Results There was no statistically significant difference in the results of the sacral base pressure test between the types of sacroiliac joint dysfunction. In terms of the results of validity, the sacral base pressure test was useful in identifying positive values of sacroiliac joint dysfunction. It was fairly helpful in correctly diagnosing patients with negative test results; however, it had only a “slight” agreement with the diagnosis for κ interpretation. Conclusions In this study, the sacral base pressure test was not a valid test for determining the presence of sacroiliac joint dysfunction or the type of dysfunction present. Further research comparing the agreement of the sacral base pressure test or other sacroiliac joint dysfunction tests with a criterion standard of diagnosis is necessary. PMID:19674694

  8. 30 CFR 7.103 - Safety system control test.

    Science.gov (United States)

    2010-07-01

    ... Areas of Underground Coal Mines Where Permissible Electric Equipment is Required § 7.103 Safety system... operate immediately when activated and stop the engine within 15 seconds. (6) The total intake air inlet...

  9. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs

  10. Advantages of the experimental animal hollow organ mechanical testing system for the rat colon rupture pressure test.

    Science.gov (United States)

    Ji, Chengdong; Guo, Xuan; Li, Zhen; Qian, Shuwen; Zheng, Feng; Qin, Haiqing

    2013-01-01

    Many studies have been conducted on colorectal anastomotic leakage to reduce the incidence of anastomotic leakage. However, how to precisely determine if the bowel can withstand the pressure of a colorectal anastomosis experiment, which is called anastomotic bursting pressure, has not been determined. A task force developed the experimental animal hollow organ mechanical testing system to provide precise measurement of the maximum pressure that an anastomotic colon can withstand, and to compare it with the commonly used method such as the mercury and air bag pressure manometer in a rat colon rupture pressure test. Forty-five male Sprague-Dawley rats were randomly divided into the manual ball manometry (H) group, the tracing machine manometry pressure gauge head (MP) group, and the experimental animal hollow organ mechanical testing system (ME) group. The rats in each group were subjected to a cut colon rupture pressure test after injecting anesthesia in the tail vein. Colonic end-to-end anastomosis was performed, and the rats were rested for 1 week before anastomotic bursting pressure was determined by one of the three methods. No differences were observed between the normal colon rupture pressure and colonic anastomotic bursting pressure, which were determined using the three manometry methods. However, several advantages, such as reduction in errors, were identified in the ME group. Different types of manometry methods can be applied to the normal rat colon, but the colonic anastomotic bursting pressure test using the experimental animal hollow organ mechanical testing system is superior to traditional methods. Copyright © 2013 Surgical Associates Ltd. Published by Elsevier Ltd. All rights reserved.

  11. Low Pressure Circuit Control and adjust System Test

    International Nuclear Information System (INIS)

    Rubio, R.O; Brendstrup, C.J; Ocampo, A.C

    2000-01-01

    The hydraulic mechanism (MSAC) is a system that will be employed in the movement of the control rods of the CAREM-25 reactor.In this report, the experimental work on a prototype of MSAC in a low pressure circuit is presented: also the methodology and conclusions.Basic thermalhydraulic data from the MSAC was obtained, and the most relevant control parameters were determined.The response of the mechanism to changes in the control parameters was also evaluated. In conclusion, the response of the MSAC fulfills the aspects of reliability and repetitive movement with water flow pulses control, in the low pressure circuit at the Laboratorio de Mecanica, Materiales y Mediciones of INVAP S.E

  12. Measurement error in pressure-decay leak testing

    International Nuclear Information System (INIS)

    Robinson, J.N.

    1979-04-01

    The effect of measurement error in presssure-decay leak testing is considered, and examples are presented to demonstrate how it can be properly accomodated in analyzing data from such tests. Suggestions for more effective specification and conduct of leak tests are presented

  13. Safety of the pressure vessels of water reactors. Prevention of sudden failure

    International Nuclear Information System (INIS)

    Petrequin, P.; Barrachin, B.

    1975-01-01

    From the safety view point the primary circuit is considered as the essential barrier against the diffusion of radioactive products in the event of fuel element failure. The safety of the vessel itself, the failure of which is not accounted for in accident analyses, is based chiefly on a series of preventive measures such as the suitable choice of materials and manufacturing process, compliances with detailed specifications concerning tests and defect tolerances, supervision in service. All these points are examined in detail when the safety analysis is performed. In this context the Service de Recherches Metallurgiques Appliquees assists the Department de Surete Nucleaire in the study of special problems such as the prevention of sudden failure and the characterisation of steels as a function of working conditions, particularly neutron irradiation. The report is thus devoted mainly to the presentation of methods to prevent sudden failure, with special emphasis on the limits of application. Some results obtained at the Service de Recherches Metallurgiques Appliquees on steels typical of those used for water reactor vessels (A533 and A508Cl.3) are given by way of example. Part two concentrates on the role of various factors influencing embrittlement by irradiation [fr

  14. Testing for altruism and social pressure in charitable giving.

    Science.gov (United States)

    DellaVigna, Stefano; List, John A; Malmendier, Ulrike

    2012-01-01

    Every year, 90% of Americans give money to charities. Is such generosity necessarily welfare enhancing for the giver? We present a theoretical framework that distinguishes two types of motivation: individuals like to give, for example, due to altruism or warm glow, and individuals would rather not give but dislike saying no, for example, due to social pressure. We design a door-to-door fund-raiser in which some households are informed about the exact time of solicitation with a flyer on their doorknobs. Thus, they can seek or avoid the fund-raiser. We find that the flyer reduces the share of households opening the door by 9% to 25% and, if the flyer allows checking a Do Not Disturb box, reduces giving by 28% to 42%. The latter decrease is concentrated among donations smaller than $10. These findings suggest that social pressure is an important determinant of door-to-door giving. Combining data from this and a complementary field experiment, we structurally estimate the model. The estimated social pressure cost of saying no to a solicitor is $3.80 for an in-state charity and $1.40 for an out-of-state charity. Our welfare calculations suggest that our door-to-door fund-raising campaigns on average lower the utility of the potential donors.

  15. Generic test platform for representative tests of safety I/C systems - 15546

    International Nuclear Information System (INIS)

    Fourestie, B.; Kuck, H.; Richter, J.; Rieche, S.; Waitz, M.

    2015-01-01

    In compliance with the IEC 61513 safety Instrumentation and Control (I/C) systems must be successfully validated in their final configuration prior to installation on site and commissioning. However the contingent need for modifications during system validation activities or subsequently during the commissioning phase may entail long and costly re-engineering of the I/C systems. With the view to ease these possible modifications, a Generic Test Platform has been developed by AREVA which allows combining a real I/C system subpart with an emulation server. This platform provides a faithful representation of the I/C System allowing crediting the validation test results carried out on this platform. (authors)

  16. Mark II containment 1/6-scale pressure suppression test program: data report no. 2

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Okazaki, Motoaki; Namatame, Ken; Shiba, Masayoshi

    1979-08-01

    This report documents experimental data from the first test phase of the Mark II Containment 1/6-Scale Pressure Suppression Test. The 1/6-Scale Test was initiated in December, 1976, to investigate the thermohydraulic responses of a BWR Mark II pressure suppression system to a postulated loss-of-coolant accident (LOCA), by means of scale model experiments. From January to June, 1977, a series of tests were performed for the Japanese BWR Owners' Group. These tests consisted of eight air-blowdown pool swell tests, three steam-blowdown pool swell tests, and twelve steam condensation tests. The dynamic responses of pressure and pool water level during the blowdown, pressure oscillation and chugging phenomena associated with unsteady condensation of steam were measured. (author)

  17. The 900 MWe water pressurized reactor safety re-examination at the occasion of their third decennial inspection

    International Nuclear Information System (INIS)

    2009-01-01

    This document reports the safety re-examination actions performed on the French 900 MWe water pressurized reactors. This process includes three stages. The first one is an inventory of safety, design and operation requirements which are defined or specified in different texts: regulations, rules, criteria and specifications. This leads to compliance studies with respect to these documents and by in situ inspections, and then to corrective recommendations. After presenting this process, the report deals with specific safety studies which are related to external or internal aggressions (fire, explosions, flooding, climate, seism), to accidental situations (primary circuit cold overpressure, severe accidents, containment, level 1 and 2 safety probabilistic studies, passive failure of safeguard circuits, vapour generator tube failure, and so on), to design and sizing of civil engineering works and systems (radioactivity measurement system, safety injection system, recirculation function liability, liability of the irradiated fuel deactivation pool cooling system)

  18. Incident at university research facility - pressure testing of gas hydrate cell

    DEFF Research Database (Denmark)

    Jensen, Niels; Jørgensen, Sten Bay

    2014-01-01

    A master student designed a cell for observing the development of gas hydrates as conditions in the cell were changed. The supervisor asked for a pressure test of the cell before the experiments started. The student chose-to perform the pressure test using compressed air and this resulted in one...

  19. Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

    2009-01-01

    Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

  20. Impact of proof test interval and coverage on probability of failure of safety instrumented function

    International Nuclear Information System (INIS)

    Jin, Jianghong; Pang, Lei; Hu, Bin; Wang, Xiaodong

    2016-01-01

    Highlights: • Introduction of proof test coverage makes the calculation of the probability of failure for SIF more accurate. • The probability of failure undetected by proof test is independently defined as P TIF and calculated. • P TIF is quantified using reliability block diagram and simple formula of PFD avg . • Improving proof test coverage and adopting reasonable test period can reduce the probability of failure for SIF. - Abstract: Imperfection of proof test can result in the safety function failure of safety instrumented system (SIS) at any time in its life period. IEC61508 and other references ignored or only elementarily analyzed the imperfection of proof test. In order to further study the impact of the imperfection of proof test on the probability of failure for safety instrumented function (SIF), the necessity of proof test and influence of its imperfection on system performance was first analyzed theoretically. The probability of failure for safety instrumented function resulted from the imperfection of proof test was defined as probability of test independent failures (P TIF ), and P TIF was separately calculated by introducing proof test coverage and adopting reliability block diagram, with reference to the simplified calculation formula of average probability of failure on demand (PFD avg ). Research results show that: the shorter proof test period and the higher proof test coverage indicate the smaller probability of failure for safety instrumented function. The probability of failure for safety instrumented function which is calculated by introducing proof test coverage will be more accurate.

  1. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    International Nuclear Information System (INIS)

    Fabry, A.

    1997-01-01

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond

  2. Nuclear reactor pressure vessel integrity insurance by crack arrestability evaluation using load from CVN tests

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.

    1997-10-15

    The present work is undertaken in the framework of nuclear reactor pressure vessel (RPV) surveillance and aims at revisiting the crack arrest approach to structural integrity insurance. This approach, performed under normal plant operation conditions, can also offer an attractive alternative to the crack initiation philosophy promoted for accidental analysis. To this end, an accidental conservative, cost effective and robust methodology is forwarded and demonstrated: it makes use of the crack arrest information contained in the instrumented Charpy V-notch impact test and/or in the shear fracture appearance of broken samples. Particular attention is paid to the appraisal of uncertainties and the related safety margin. The resulting capability is placed in perspective with the state-of-the-art crack initiation methodology based on the slow bend testing of recracked specimens, presently under standardization world-wide. The investigation leads to highlight three conceptual weaknesses of current enfgineering and regulatory practices. Improved crack arrestability evaluation emerges as an optimal approach to insure safe PWR operation up to design end-of-life and beyond.

  3. An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

    International Nuclear Information System (INIS)

    Waltar, A.E.; Padilla, A. Jr.

    1990-11-01

    The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs

  4. Ultrasonic testing of installed low-pressure turbine shafts

    International Nuclear Information System (INIS)

    Hildmann, I.; Voelker, J.; Ewald, J.

    1987-01-01

    Transverse defects in the admission area of double-flow LP turbine shafts with shrink-on wheel disks can be detected during the onset of crack growth by means of a newly developed test concept with slightly oblique longitudinal US wave incidence, and crack size estimates can be made. For process development and system adjustment a large reference specimen with circular and circular segment-type test reflectors was used. The results of comparative measurements with different types of devices and probes of different transducer size, test frequency and pulse length are presented, and the choice of the technical testing details is substantiated. (orig./DG) [de

  5. 78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Science.gov (United States)

    2013-05-01

    ... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

  6. Sled Tests Using the Hybrid III Rail Safety ATD and Workstation Tables for Passenger Trains

    Science.gov (United States)

    2017-08-01

    The Hybrid III Rail Safety (H3-RS) anthropomorphic test device (ATD) is a crash test dummy developed in the United Kingdom to evaluate abdomen and lower thorax injuries that occur when passengers impact workstation tables during train accidents. The ...

  7. Hydraulic pressure pulses with elastic and plastic structural flexibility: test and analysis (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schwirian, R.E.

    1978-03-01

    Pressure pulse tests were conducted with a flexible test section in a test vessel filled with room temperature water. The pressure pulses were generated with a drop hammer and piston pulse generator and were of a sufficient magnitude to cause plastic deformation of the test section. Because of the strong pressure relief effect of the deforming test section, pressure pulse magnitudes were below 265 psig in magnitude and had durations of 50 to 55 msecs. Calculations performed with the FLASH-35 bi-linear hysteresis model of structural deformation show good agreement with experiment. In particular, FLASH 35 adequately predicts the decrease in peak pressure and the increase in pulse duration due to elastic and plastic deformation of the test section. Predictions of flexible member motion are good, but are less satisfactory than the pressure pulse results due to uncertainties in the values of yield point and beyond yield stiffness used to model the various flexible members. Coupled with this is a strong sensitivity of the FLASH 35 predictions to the values of yield point and beyond yield stiffness chosen for the various flexible members. The test data versus calculation comparisons presented here provide preliminary qualification for FLASH 35 calculations of transient hydraulic pressures and pressure differentials in the presence of flexible structural members which deform both elastically and plastically

  8. High temperature helium test rig with prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schmidl, H.

    1975-10-01

    The report gives a short description of the joint project prestressed concrete vessel-helium test station as there is the building up of the concrete structure, the system of instrumentation, the data processing, the development of the helium components as well as the testing programs. (author)

  9. 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)

  10. Environmental testing of a prototypic digital safety channel, Phase I: System design and test methodology

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Turner, G.W.; Mullens, J.A. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    A microprocessor-based reactor trip channel has been assembled for environmental testing under an Instrumentation and Control (I&C) Qualification Program sponsored by the US Nuclear Regulatory Commission. The goal of this program is to establish the technical basis and acceptance criteria for the qualification of advanced I&C systems. The trip channel implemented for this study employs technologies and digital subsystems representative of those proposed for use in some advanced light-water reactors (ALWRs) such as the Simplified Boiling Water Reactor (SBWR). It is expected that these tests will reveal any potential system vulnerabilities for technologies representative of those proposed for use in ALWRs. The experimental channel will be purposely stressed considerably beyond what it is likely to experience in a normal nuclear power plant environment, so that the tests can uncover the worst-case failure modes (i.e., failures that are likely to prevent an entire trip system from performing its safety function when required to do so). Based on information obtained from this study, it may be possible to recommend tests that are likely to indicate the presence of such failure mechanisms. Such recommendations would be helpful in augmenting current qualification guidelines.

  11. Environmental testing of a prototypic digital safety channel, phase I: System design and test methodology

    International Nuclear Information System (INIS)

    Korsah, K.; Turner, G.W.; Mullens, J.A.

    1995-01-01

    A microprocessor-based reactor trip channel has been assembled for environmental testing under an Instrumentation and Control (I ampersand C) Qualification Program sponsored by the U.S. Nuclear Regulatory Commission. The goal of this program is to establish the technical basis for the qualification of advanced I ampersand C systems. The trip channel implemented for this study employs technologies and digital subsystems representative of those proposed for use in some advanced light-water reactors (ALNWS) such as the Simplified Boiling Water Reactor (SBNW) and AP600. It is expected that these tests will reveal any potential system vulnerabilities for technologies representative of those proposed for use in ALNWS. The experimental channel will be purposely stressed considerably beyond what it is likely to experience in a normal nuclear power plant environment, so that the tests can uncover the worst-case failure modes (i.e., failures that are likely to prevent an entire trip system from performing its safety function when required to do so). Based on information obtained from this study, it may be possible to recommend tests that are likely to indicate the presence of such failure mechanisms. Such recommendations would be helpful in augmenting current qualification guidelines

  12. A study on implementation of dynamic safety system in programmable logic controller for pressurized water reactor

    International Nuclear Information System (INIS)

    Kim, Ung Soo

    1997-02-01

    The dynamic safety system (DSS) is a computer based reactor protection system that has dynamic self-testing feature and fail-safe nature inherently. The inherent dynamic self-testing feature and fail-safe design provide a high level of reliability and low spurious trip rate. We can also reduce the time and human efforts to maintain the system by virtue of those features. Therefore, the application of the DSS to PWR has many advantages. The DSS has been applied only to advanced gas-cooled reactor (AGR) in the UK. In order to apply the DSS for PWR, the DSS has to be modified because there exist many differences between PWR and AGR for which the DSS was tested and installed. These differences are trip algorithms, monitored parameters, trip logics, and other conditions. In this study, the DSS algorithm is modified for PWR first. The modified DSS has several new features : 1) The modified DSS tests and processes time-dependent parameters, while the original DSS does not. 2) It has flexibility for handling several types of voting logic but the original DSS handles the only one type of voting - 2 out of 4 coincidence logic. Then, in this study, the modified DSS is implemented in programmable logic controller (PLC) using the ladder logic. Finally, the modified DSS is tested in two ways in this work : 1) The manual test is performed using direct input through the human computer interface (HCI) system. 2) The scenario based test is performed using input from the FISA-2/WS simulator. From the test results, it is shown that the modified DSS operates correctly in all conditions

  13. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  14. Fast flux test facility final safety analysis report amendment 79

    International Nuclear Information System (INIS)

    Dautel, W.A.

    1999-01-01

    This document is provided to replace, remove, or add applicable pages to the chapters on: Heat Transport System; Containment and Structures; Auxiliary Systems; Reactor Refueling System; Conduct of Operations; Safety Analysis; Quality Assurance; FFTF Criticality Specifications; and Appendix H's TRIGA Fuel Storage System

  15. Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

    International Nuclear Information System (INIS)

    Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

    2002-01-01

    A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

  16. Safety assessment for the 118-B-1 Burial Ground excavation treatability tests. Revision 2

    International Nuclear Information System (INIS)

    Zimmer, J.J.; Frain, J.M.

    1994-12-01

    This revision of the Safety Assessment provides an auditable safety analysis of the hazards for the proposed treatability test activities per DOE-EM-STD-5502-94, DOE Limited Standard, Hazard Baseline Documentation (DOE 1994). The proposed activities are classified as radiological activities and as such, no longer require Operational Safety Limits (OSLs). The OSLS, Prudent Actions, and Institutional and Organization Controls have been removed from this revision and replaced with ''Administrative Actions Important to Safety,'' as determined by the hazards analysis. Those Administrative Actions Important to Safety are summarized in Section 1.1, ''Assessment Summary.''

  17. Occupational safety and health textbook for radiological personnel employed in structural material testing

    International Nuclear Information System (INIS)

    Abraham, J.

    1981-01-01

    The comprehensive textbook for X-ray and radiological testing personnel includes requirements and rules of occupational safety and health on the basis of Hungarian and international (mainly German) literature. In the chapter Fundamentals, X-ray and radioactive radiations, their measurements and biological effects, doses etc are described. In the chapter Occupational safety and health, the jobs representing radiation hazards are listed and safety regulations for them are reported. Finally, information for prevention and first aid is presented. Control questions are added to each part. The Appendix contains safety standards and regulations, information on legal aspects of safety and radiation protection as well as recommendations. (Sz.J.)

  18. Split-Hopkinson Pressure Bar: an experimental technique for high strain rate tests

    International Nuclear Information System (INIS)

    Sharma, S.; Chavan, V.M.; Agrawal, R.G.; Patel, R.J.; Kapoor, R.; Chakravartty, J.K.

    2011-06-01

    Mechanical properties of materials are, in general, strain rate dependent, i.e. they respond differently at quasi-static and higher strain rate condition. The Split-Hopkinson Pressure Bar (SHPB), also referred to as Kolsky bar is a commonly used setup for high strain rate testing. SHPB is suitable for high strain rate test in strain rate range of 10 2 to 10 4 s -1 . These high strain rate data are required for safety and structural integrity assessment of structures subjected to dynamic loading. As high strain rate data are not easily available in open literature need was felt for setting up such high strain rate testing machine. SHPB at BARC was designed and set-up inhouse jointly by Refuelling Technology Division and Mechanical Metallurgy Division, at Hall no. 3, BARC. A number of conceptual designs for SHPB were thought of and the optimized design was worked out. The challenges of precision tolerance, straightness in bars and design and proper functioning of pneumatic gun were met. This setup has been used extensively to study the high strain rate material behavior. This report introduces the SHPB in general and the setup at BARC in particular. The history of development of SHPB, the basic formulations of one dimensional wave propagation, the relations between the wave velocity, particle velocity and elastic strain in a one dimensional bar, and the equations used to obtain the final stress vs. strain curves are described. The calibration of the present setup, the pre-test calculations and the posttest analysis of data are described. Finally some of the experimental results on different materials such as Cu, SS305, SA516 and Zr, at room temperature and elevated temperatures are presented. (author)

  19. Need and trends of volumetric tests in recurring inspection of pressurized components in pressurized water reactors

    International Nuclear Information System (INIS)

    Bergemann, W.

    1982-01-01

    On the basis of the types of stress occurring in nuclear power plants and of practical results it has been shown that cracks in primary circuit components arise due to operating stresses in both the materials surfaces and the bulk of the materials. For this reason, volumetric materials testing is necessary in addition to surface testing. An outlook is given on the trends of volumetric testing. (author)

  20. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    International Nuclear Information System (INIS)

    Nurkkala, P.; Hoikkanen, J.

    1997-01-01

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. 'grounded' and 'with goose neck'). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.)

  1. Measuring the initial earth pressure of granite using hydraulic fracturing test; Goseong and Yuseong areas

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byoung Yoon; Bae, Dae Seok; Kim, Chun Soo; Kim, Kyung Su; Koh, Young Kwon; Won, Kyung Sik [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-02-01

    This report provides the initial earth pressure of granitic rocks obtained from Deep Core Drilling Program which is carried out as part of the assessment of deep geological environmental condition. These data are obtained by hydraulic fracturing test in three boreholes drilled up to 350{approx}500 m depth at the Yuseong and Goseong sites. These sites were selected based on the result of preliminary site evaluation study. The boreholes are NX-size (76 mm) and vertical. The procedure of hydraulic fracturing test is as follows: - Selecting the testing positions by preliminary investigation using BHTV logging. - Performing the hydraulic fracturing test at each selected position with depth.- Estimating the shut-in pressure by the bilinear pressure-decay-rate method. - Estimating the fracture reopening pressure from the pressure-time curves.- Estimating the horizontal principal stresses and the direction of principal stresses. 65 refs., 39 figs., 12 tabs. (Author)

  2. New vision of the control organisms in industrial safety and maintenance, based approach to new pressure equipment

    International Nuclear Information System (INIS)

    Bernardez Garcia, A.

    2010-01-01

    Control agencies are companies dedicated to the verification of compliance with the safety of products and facilities as administrative regulation in industrial safety through certification activities, testing, inspection or audit.Changes have been made that will stimulate the increase of companies engaged in this sector.

  3. Split-Hopkinson pressure bar tests on pure tantalum

    International Nuclear Information System (INIS)

    Dick, Richard D.; Armstrong, Ronald W.; Williams, John D.

    1998-01-01

    Pure tantalum (Ta) was loaded in compression by a split-Hopkinson pressure bar (SHPB) to strain rates from 450 to 6350 s -1 . The results are compared with SHPB data for commercial Ta and with predictions from the constitutive model for Ta developed by Zerilli and Armstrong (Z-A). The main conclusions are: (1) the flow stress versus log strain rate agree with the Z-A constitutive model and other reported data, (2) uniform strain exponents computed on a true stress-strain basis for pure Ta are somewhat greater than those determined from SHPB data for commercial Ta, and (3) in both cases the uniform strain exponents versus log strain rate are in good agreement with predictions from the Z-A constitutive model for strain rates above 1500 s -1 without a clear indication of dislocation generation

  4. 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)

  5. DELPHIN - a new system for testing reactor pressure vessels

    International Nuclear Information System (INIS)

    Dressler, K.

    1999-01-01

    The author discusses the question of whether a new concept for testing RPVs is necessary. He concentrates his exposition upon the RPV testing system DELPHIN recently developed by ABB, which has been successfully employed in vessel-interior tests since 1998. The new system reduces both the time required and the financial costs for RPV tests. The tests have become more efficient particularly as a result of new developments in the fields of handling machinery and microelectronics. As an example of the improved quality, the author quotes the ultrasonic system ZAQUS: thanks to the high quality of the ultrasonic data, rapid comparison with the results of earlier repreated tests on the RPV is now possible. Since problems of interpretation did not arise, the overall results in an initial application were available only two hours after the last data recording. The author's verdict: DELPHIN has successfully undergone its 'baptism by fire'; still in need of improvement, he states, is the occupany time of the pond, which is not yet as short as targeted. (orig.) [de

  6. Development of a test rig and its application for validation and reliability testing of safety-critical software

    Energy Technology Data Exchange (ETDEWEB)

    Thai, N D; McDonald, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    This paper describes a versatile test rig developed by AECL for functional testing of safety-critical software used in the process trip computers of the Wolsong CANDU stations. The description covers the hardware and software aspects of the test rig, the test language and its interpreter, and other major testing software utilities such as the test oracle, sampler and profiler. The paper also discusses the application of the rig in the final stages of testing of the process trip computer software, namely validation and reliability tests. It shows how random test cases are generated, test scripts prepared and automatically run on the test rig. The versatility of the rig is further demonstrated in other types of testing such as sub-system tests, verification of the test oracle, testing of newly-developed test script, self-test and calibration. (author). 5 tabs., 10 figs.

  7. Development of a test rig and its application for validation and reliability testing of safety-critical software

    International Nuclear Information System (INIS)

    Thai, N.D.; McDonald, A.M.

    1995-01-01

    This paper describes a versatile test rig developed by AECL for functional testing of safety-critical software used in the process trip computers of the Wolsong CANDU stations. The description covers the hardware and software aspects of the test rig, the test language and its interpreter, and other major testing software utilities such as the test oracle, sampler and profiler. The paper also discusses the application of the rig in the final stages of testing of the process trip computer software, namely validation and reliability tests. It shows how random test cases are generated, test scripts prepared and automatically run on the test rig. The versatility of the rig is further demonstrated in other types of testing such as sub-system tests, verification of the test oracle, testing of newly-developed test script, self-test and calibration. (author). 5 tabs., 10 figs

  8. Debris filtering effectiveness and pressure drop tests of debris resistance-bottom end piece

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Song, Chul Hwa; Chung, Heung June; Won, Soon Yeun; Cho, Young Ro; Kim, Bok Deuk

    1992-03-01

    In this final report, described are the test conditions and test procedures for the debris filtering effectiveness and pressure drop tests for developing the Debris Resistance-Bottom End Piece (DR-BEP). And the test results are tabulated for later evaluation. (Author)

  9. Comparison of under-pressure and over-pressure pulse tests conducted in low-permeability basalt horizons at the Hanford Site, Washington State

    International Nuclear Information System (INIS)

    Thorne, P.D.; Spane, F.A. Jr.

    1984-10-01

    Over-pressure pulse tests (pressurized slug tests have been widely used by others for hydraulic characterization of low-permeability ( -8 m/sec) rock formations. Recent field studies of low-permeability basalt horizons at the Hanford Site, Washington, indicate that the under-pressure pulse technique is also a viable test method for hydraulic characterization studies. For over-pressure pulse tests, fluid within the test system is rapidly pressurized and the associated pressure decay is monitored as compressed fluid within the test system expands and flows into the test formation. Under-pressure pulse tests are conducted in a similar manner by abruptly decreasing the pressure of fluid within the test system, and monitoring the associated increase in pressure as fluid flows from the formation into the test system. Both pulse test methods have been used in conjunction with other types of tests to determine the hydraulic properties of selected low-permeability basalt horizons at Hanford test sites. Results from both pulse test methods generally provide comparable estimates of hydraulic properties and are in good agreement with those from other tests

  10. General-Purpose Heat Source development: Safety Verification Test Program. Bullet/fragment test series

    Energy Technology Data Exchange (ETDEWEB)

    George, T.G.; Tate, R.E.; Axler, K.M.

    1985-05-01

    The radioisotope thermoelectric generator (RTG) that will provide power for space missions contains 18 General-Purpose Heat Source (GPHS) modules. Each module contains four /sup 238/PuO/sub 2/-fueled clads and generates 250 W/sub (t)/. Because a launch-pad or post-launch explosion is always possible, we need to determine the ability of GPHS fueled clads within a module to survive fragment impact. The bullet/fragment test series, part of the Safety Verification Test Plan, was designed to provide information on clad response to impact by a compact, high-energy, aluminum-alloy fragment and to establish a threshold value of fragment energy required to breach the iridium cladding. Test results show that a velocity of 555 m/s (1820 ft/s) with an 18-g bullet is at or near the threshold value of fragment velocity that will cause a clad breach. Results also show that an exothermic Ir/Al reaction occurs if aluminum and hot iridium are in contact, a contact that is possible and most damaging to the clad within a narrow velocity range. The observed reactions between the iridium and the aluminum were studied in the laboratory and are reported in the Appendix.

  11. General-Purpose Heat Source Safety Verification Test program: Edge-on flyer plate tests

    International Nuclear Information System (INIS)

    George, T.G.

    1987-03-01

    The radioisotope thermoelectric generator (RTG) that will supply power for the Galileo and Ulysses space missions contains 18 General-Purpose Heat Source (GPHS) modules. The GPHS modules provide power by transmitting the heat of 238 Pu α-decay to an array of thermoelectric elements. Each module contains four 238 PuO 2 -fueled clads and generates 250 W(t). Because the possibility of a launch vehicle explosion always exists, and because such an explosion could generate a field of high-energy fragments, the fueled clads within each GPHS module must survive fragment impact. The edge-on flyer plate tests were included in the Safety Verification Test series to provide information on the module/clad response to the impact of high-energy plate fragments. The test results indicate that the edge-on impact of a 3.2-mm-thick, aluminum-alloy (2219-T87) plate traveling at 915 m/s causes the complete release of fuel from capsules contained within a bare GPHS module, and that the threshold velocity sufficient to cause the breach of a bare, simulant-fueled clad impacted by a 3.5-mm-thick, aluminum-alloy (5052-T0) plate is approximately 140 m/s

  12. Explosion overpressure test series: General-Purpose Heat Source development: Safety Verification Test program

    International Nuclear Information System (INIS)

    Cull, T.A.; George, T.G.; Pavone, D.

    1986-09-01

    The General-Purpose Heat Source (GPHS) is a modular, radioisotope heat source that will be used in radioisotope thermoelectric generators (RTGs) to supply electric power for space missions. The first two uses will be the NASA Galileo and the ESA Ulysses missions. The RTG for these missions will contain 18 GPHS modules, each of which contains four 238 PuO 2 -fueled clads and generates 250 W/sub (t)/. A series of Safety Verification Tests (SVTs) was conducted to assess the ability of the GPHS modules to contain the plutonia in accident environments. Because a launch pad or postlaunch explosion of the Space Transportation System vehicle (space shuttle) is a conceivable accident, the SVT plan included a series of tests that simulated the overpressure exposure the RTG and GPHS modules could experience in such an event. Results of these tests, in which we used depleted UO 2 as a fuel simulant, suggest that exposure to overpressures as high as 15.2 MPa (2200 psi), without subsequent impact, does not result in a release of fuel

  13. Feasibility and Safety of Pressurized Intraperitoneal Aerosol Chemotherapy for Peritoneal Carcinomatosis: A Retrospective Cohort Study

    Directory of Open Access Journals (Sweden)

    Martin Hübner

    2017-01-01

    Full Text Available Background. Pressurized intraperitoneal aerosol chemotherapy (PIPAC has been introduced as a novel repeatable treatment for peritoneal carcinomatosis. The available evidence from the pioneer center suggests good tolerance and high response rates, but independent confirmation is needed. A single-center cohort was analyzed one year after implementation for feasibility and safety. Methods. PIPAC was started in January 2015, and every patient was entered into a prospective database. This retrospective analysis included all consecutive patients operated until April 2016 with emphasis on surgical feasibility and early postoperative outcomes. Results. Forty-two patients (M : F = 8 : 34, median age 66 (59–73 years with 91 PIPAC procedures in total (4×: 1,  3×: 17,  2×: 12, and  1×: 12 were analyzed. Abdominal accessibility rate was 95% (42/44; laparoscopic access was not feasible in 2 patients with previous HIPEC. Median initial peritoneal carcinomatosis index (PCI was 10 (IQR 5–17. Median operation time was 94 min (89–108 with no learning curve observed. One PIPAC application was postponed due to intraoperative intestinal lesion. Overall morbidity was 9% with 7 minor complications (Clavien I-II and one PIPAC-unrelated postoperative mortality. Median postoperative hospital stay was 3 days (2-3. Conclusion. Repetitive PIPAC is feasible in most patients with refractory carcinomatosis of various origins. Intraoperative complications and postoperative morbidity rates were low. This encourages prospective studies assessing oncological efficacy.

  14. Licensing assessment of the CANDU pressurized heavy water reactor. Volume I. Preliminary safety information document

    International Nuclear Information System (INIS)

    1977-06-01

    The PHWR design contains certain features that will require significant modifications to comply with USNRC siting and safety requirements. The most significant of these features are the reactor vessel; control systems; quality assurance program requirements; seismic design of structures, systems and components; and providing an inservice inspection program capability. None of these areas appear insolvable with current state-of-the-art engineering or with upgrading of the quality assurance program for components constructed outside of the USA. In order to be licensed in the U. S., the entire reactor assembly would have to be redesigned to comply with ASME Boiler and Pressure Vessel Code, Section III, Division 1 and Division 2. A summary matrix at the end of this volume identifies compliance of the systems and structures of the PHWR plant with the USNRC General Design Criteria. The matrix further identifies the estimated incremental cost to a 600 MWe PHWR that would be required to license the plant in the U. S. Further, the matrix identifies whether or not the incremental licensing cost is size dependent and the relative percentage of the base direct cost of a Canadian sited plant

  15. Tensile and burst tests in support of the cadmium safety rod failure evaluation

    International Nuclear Information System (INIS)

    Thomas, J.K.

    1992-02-01

    The reactor safety rods may be subjected to high temperatures due to gamma heating after the core coolant level has dropped during the ECS phase of hypothetical LOCA event. Accordingly, an experimental safety rod testing subtask was established as part of a task to address the response of reactor core components to this accident. This report discusses confirmatory separate effects tests conducted to support the evaluation of failures observed in the safety rod thermal tests. As part of the failure evaluation, the potential for liquid metal embrittlement (LME) of the safety rod cladding by cadmium (Cd) -- aluminum (Al) solutions was examined. Based on the test conditions, literature data, and U-Bend tests, its was concluded that the SS304 safety rod cladding would not be subject to LME by liquid Cd-Al solutions under conditions relevant to the safety rod thermal tests or gamma heating accident. To confirm this conclusion, tensile tests on SS304 specimens were performed in both air and liquid Cd-Al solutions with the range of strain rates, temperatures, and loading conditions spanning the range relevant to the safety rod thermal tests and gamma heating accident

  16. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  17. Test report: Preliminary tests for the High Flux Reactor: Experimental determination of flow redistribution conditions at pressures between 4 and 5 kg/cm2 abs in a rectangular channel 2 mm thick and 60 cm long

    International Nuclear Information System (INIS)

    Schleisiek, K.; Dumaine, J.C.

    1989-01-01

    In the context of safety research for the OSIRIS reactor, tests have been performed on the Super BOB cell with a view to determining experimentally the internal characteristics (or ''S'' curves) of a channel with a rectangular heating cross-section 2 x 38 mm and 600 mm long. During these tests the maximum pressure at the channel exit was brought to 3 kg/cm 2 abs. The pressurization level in the High Flux Reactor will be higher. That is why tests have been carried out at maximum pressure of 5 kg/cm 2 abs allowable on the ''super BOB'' loop without modifying it. The first objective of this test series was to determine the ''S'' curves and the exchange coefficients experimentally. This document discusses the test conditions and test results

  18. Fluid dispersal from safety cannulas: an in vitro comparative test.

    Science.gov (United States)

    Rosenthal, Victor D; Hughes, Gavin

    2015-03-01

    We report a comparative laboratory study between 2 peripheral intravenous catheters equipped with a passive fully automatic safety mechanism to assess generation of blood droplets during withdrawal. One presented no fluid droplets, whereas the other presented droplets in 48% and 60% for the best and worst case, with analysis of variance showing positive effects on the number of droplets generated (P blood splatter. Copyright © 2015 Association for Professionals in Infection Control and Epidemiology, Inc. Published by Elsevier Inc. All rights reserved.

  19. An Ultra-low Frequency Modal Testing Suspension System for High Precision Air Pressure Control

    Directory of Open Access Journals (Sweden)

    Qiaoling YUAN

    2014-05-01

    Full Text Available As a resolution for air pressure control challenges in ultra-low frequency modal testing suspension systems, an incremental PID control algorithm with dead band is applied to achieve high-precision pressure control. We also develop a set of independent hardware and software systems for high-precision pressure control solutions. Taking control system versatility, scalability, reliability, and other aspects into considerations, a two-level communication employing Ethernet and CAN bus, is adopted to complete such tasks as data exchange between the IPC, the main board and the control board ,and the pressure control. Furthermore, we build a single set of ultra-low frequency modal testing suspension system and complete pressure control experiments, which achieve the desired results and thus confirm that the high-precision pressure control subsystem is reasonable and reliable.

  20. Area Safety Program for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Rappe, G.M.

    1984-10-01

    Overall the Area Safety Program has proved to be a very successful operation. There is no doubt that a safety program organized through line management is the best way to involve all personnel. Naturally, when the program was first started, there was some criticism and a certain resistance on the part of a few individuals to fully participate. However, once the program was underway and it could be seen that it was working to everyone's advantage, this reluctance disappeared and a spirit of full cooperation is now enjoyed. It is very important that for this success to continue there must be a two way flow of information, both from the Area Safety Coordinators up through line management, and from senior management, with decisions and answers, back down through the management chain with the utmost dispatch. As with all programs, there is still room for improvement. This program has started a review cycle with a view to streamlining certain areas and possibly increasing its scope in others

  1. Predictive value of lumbar infusion test in normal pressure ...

    African Journals Online (AJOL)

    Of all cases in the study, 18 (90%) had positive test results and were operated on; 16 (80%) of patients reported subjective improvement, and postoperative assessments verified the improvements in 15 patients (75%). Improvements were highly significant in walking and memory. Most of the patients improved by surgery ...

  2. Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

    International Nuclear Information System (INIS)

    2007-06-01

    be used in reassessing the safety of individual operating plants. In 1998, the IAEA completed IAEA-TECDOC-1044 entitled Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for their Resolution and established the associated LWRGSIDB database (Computer Manual Series No. 13). The present compilation, which is based on broad international experience, is an extension of this work to cover pressurized heavy water reactors (PHWRs). As in the case of LWRs, it is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It addresses generic safety issues identified in nuclear power plants using PHWRs. In most cases, the measures taken or planned to resolve these issues are also identified. The work on this report was initiated by the Senior Regulators of Countries Operating CANDU-Type Nuclear Power Plants at one of their annual meetings. It was carried out within the framework of the IAEA's programme on National Regulatory Infrastructure for Nuclear Installation Safety and serves to enhance regulatory effectiveness through the exchange of safety related information

  3. The Self-Calibration Test of flowmeter installed in STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) facility

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Minhwan; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The objective of this study is to describe the procedure of the self-calibration test for the flowmeters and to analyze the result of the test. In this work, the test procedure of the self-calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility was described and the test result was analyzed. In regard to the long-term SFR development plan, a large-scale sodium thermal-hydraulic test project is being progressed by KAERI. This project is called STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment), and it is proceeding by adopting the QA (Quality Assurance) program. Due to the specificity of an experiment using sodium(Na) categorized as Class 3(pyrophoric material and water-prohibiting substance) by the Safety Control of Dangerous Substances Act, it is necessary to apply QA in consideration of the sodium experiment environment in certain parts. The one of them is about calibration of measuring instrument such as a flowmeter, thermocouple and pressure gauge. It is described in the QAP (Quality Assurance Procedures) of KAERI that calibration work should be conducted in accordance with self-calibration procedures in a special case where conventional calibration is not practicable. The calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility is the typical example. As a result of test, it was confirmed that the flowmeters meet the pass criterion. Therefore, it was concluded that the flowmeters maintain instrument capacity a year ago.

  4. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  5. Making work safer: testing a model of social exchange and safety management.

    Science.gov (United States)

    DeJoy, David M; Della, Lindsay J; Vandenberg, Robert J; Wilson, Mark G

    2010-04-01

    This study tests a conceptual model that focuses on social exchange in the context of safety management. The model hypothesizes that supportive safety policies and programs should impact both safety climate and organizational commitment. Further, perceived organizational support is predicted to partially mediate both of these relationships. Study outcomes included traditional outcomes for both organizational commitment (e.g., withdrawal behaviors) as well as safety climate (e.g., self-reported work accidents). Questionnaire responses were obtained from 1,723 employees of a large national retailer. Using structural equation modeling (SEM) techniques, all of the model's hypothesized relationships were statistically significant and in the expected directions. The results are discussed in terms of social exchange in organizations and research on safety climate. Maximizing safety is a social-technical enterprise. Expectations related to social exchange and reciprocity figure prominently in creating a positive climate for safety within the organization. Copyright 2010 Elsevier Ltd. All rights reserved.

  6. Standards for radiation protection instrumentation: design of safety standards and testing procedures

    International Nuclear Information System (INIS)

    Meissner, Frank

    2008-01-01

    This paper describes by means of examples the role of safety standards for radiation protection and the testing and qualification procedures. The development and qualification of radiation protection instrumentation is a significant part of the work of TUV NORD SysTec, an independent expert organisation in Germany. The German Nuclear Safety Standards Commission (KTA) establishes regulations in the field of nuclear safety. The examples presented may be of importance for governments and nuclear safety authorities, for nuclear operators and for manufacturers worldwide. They demonstrate the advantage of standards in the design of radiation protection instrumentation for new power plants, in the upgrade of existing instrumentation to nuclear safety standards or in the application of safety standards to newly developed equipment. Furthermore, they show how authorities may proceed when safety standards for radiation protection instrumentation are not yet established or require actualization. (author)

  7. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Min, Sohn Jae; Kang, Y. H.; Kim, B. G. [and others

    2001-11-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO, the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT. The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. The out-of-pile test system for pressure measurement was developed, and the test with the LVDT at room temperature(19 .deg. C) were performed. A out-of-pile test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2} and repeated 6 times at each condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. This report describes the system configuration, the out-of-pile test procedures, and the results. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics for the detail design of the fuel irradiation capsule.

  8. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

  9. Effects of time pressure and noise on non-destructive testing

    Energy Technology Data Exchange (ETDEWEB)

    Enkvist, J.; Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology; Edland, A. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2001-12-01

    Manual ultrasonic testing (UT) is the most frequently used non-destructive testing (NDT) method for in-service inspection of components important to safety and/or plant availability. Earlier, great variations have been found in operator performance, often attributed to operator fatigue. However, no conclusive findings have been reported. According to the Yerkes-Dodson law there is an optimal arousal level where performance is highest, for simple tasks this optimum is higher than for more complex tasks. In the present study twenty operators performed manual ultrasonic inspections of six test pieces with manufactured flaws. The operators performed the inspections under stress (high arousal - time pressure and noise) and non-stress conditions; one condition the first day and the other the second and last day. It was hypothesised that the stress condition led to a level of arousal so high that it would affect the results negatively. The results confirmed that the operators were affected by the stress condition. However, contrary to the hypotheses it was found that the manipulation increased operator performance. Operators with the stress condition the first day performed better than the other operators did (under both the stress and the non-stress condition). This was interpreted as the 'stress first' (group 1) operators had established efficient performance patterns the first day - affecting also the second day. Operators beginning with stress condition also tended to be more motivated. It was concluded that operator performance is affected by arousal. The operators with non-stress first (group 2) worked hard with the complex task but their arousal level was assumed to be above the optimal, resulting in a low hit rate.

  10. Effects of time pressure and noise on non-destructive testing

    Energy Technology Data Exchange (ETDEWEB)

    Enkvist, J; Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology; Edland, A [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2001-12-01

    Manual ultrasonic testing (UT) is the most frequently used non-destructive testing (NDT) method for in-service inspection of components important to safety and/or plant availability. Earlier, great variations have been found in operator performance, often attributed to operator fatigue. However, no conclusive findings have been reported. According to the Yerkes-Dodson law there is an optimal arousal level where performance is highest, for simple tasks this optimum is higher than for more complex tasks. In the present study twenty operators performed manual ultrasonic inspections of six test pieces with manufactured flaws. The operators performed the inspections under stress (high arousal - time pressure and noise) and non-stress conditions; one condition the first day and the other the second and last day. It was hypothesised that the stress condition led to a level of arousal so high that it would affect the results negatively. The results confirmed that the operators were affected by the stress condition. However, contrary to the hypotheses it was found that the manipulation increased operator performance. Operators with the stress condition the first day performed better than the other operators did (under both the stress and the non-stress condition). This was interpreted as the 'stress first' (group 1) operators had established efficient performance patterns the first day - affecting also the second day. Operators beginning with stress condition also tended to be more motivated. It was concluded that operator performance is affected by arousal. The operators with non-stress first (group 2) worked hard with the complex task but their arousal level was assumed to be above the optimal, resulting in a low hit rate.

  11. Effects of time pressure and noise on non-destructive testing

    International Nuclear Information System (INIS)

    Enkvist, J.; Svenson, Ola

    2001-12-01

    Manual ultrasonic testing (UT) is the most frequently used non-destructive testing (NDT) method for in-service inspection of components important to safety and/or plant availability. Earlier, great variations have been found in operator performance, often attributed to operator fatigue. However, no conclusive findings have been reported. According to the Yerkes-Dodson law there is an optimal arousal level where performance is highest, for simple tasks this optimum is higher than for more complex tasks. In the present study twenty operators performed manual ultrasonic inspections of six test pieces with manufactured flaws. The operators performed the inspections under stress (high arousal - time pressure and noise) and non-stress conditions; one condition the first day and the other the second and last day. It was hypothesised that the stress condition led to a level of arousal so high that it would affect the results negatively. The results confirmed that the operators were affected by the stress condition. However, contrary to the hypotheses it was found that the manipulation increased operator performance. Operators with the stress condition the first day performed better than the other operators did (under both the stress and the non-stress condition). This was interpreted as the 'stress first' (group 1) operators had established efficient performance patterns the first day - affecting also the second day. Operators beginning with stress condition also tended to be more motivated. It was concluded that operator performance is affected by arousal. The operators with non-stress first (group 2) worked hard with the complex task but their arousal level was assumed to be above the optimal, resulting in a low hit rate

  12. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  13. Testing - Smart strategy for safety and mission quality

    Science.gov (United States)

    Rodney, George A.

    The paper is concerned with the need for a comprehensive test plan for the Space Station Freedom (SST) that would fully verify specification compliance and be based on an error budget. In particular, attention is given to some lessons learned from other NASA programs and the principal challenges for SSF testing, including phase C/D/E agreements, testing parameters, phase testing, and the human element. The importance of close teamwork between the NASA/Contractor systems engineers and assurance engineers is emphasized.

  14. Testing of VVER reactor pressure vessels by TOFD method

    International Nuclear Information System (INIS)

    Skala, Z.; Vit, J.

    2002-01-01

    The Time of Flight Diffraction Method (TOFD) - one of the new testing methods capable to obtain the real dimensions of flaws - is presented in the paper.The laboratory experiments on samples with artificial flaws and samples with artificially prepared cracks confirmed the high accuracy of flaw through wall extent sizing by TOFD. This accuracy was confirmed by qualification of methods and systems used by Skoda JS for the in-service inspections of WWER 440 vessel circumferential weld. The qualification also confirmed the ability of TOFD to detect reliably flaws, which can are not reliably detected by standard pulse echo testing. Based on the result of experiments and qualification, the TOFD method shall be used routinely by Skoda JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level

  15. Manufacturing and testing experience for FFTF major safety related components

    International Nuclear Information System (INIS)

    Peckinpaugh, C.L.

    1976-01-01

    Experience with FFTF Heat Transport System components during design, manufacturing, and prototype testing is dscussed. Specifically the special design features and the results of the testing performed to assure that the designs provide for safe operation are outlined. Particular emphasis is placed on the full size prototype testing programs and the valuable experience gained

  16. 23. MPA-Seminar: Safety and reliability of plant technology with special emphasis on behaviour of pressurized components and systems at increased loading. Vol. 2. Papers 27-50

    International Nuclear Information System (INIS)

    1998-01-01

    This book is dedicated to the components of nuclear and conventional power plants with special emphasis on the behaviour of pressurized components and systems. The following topics are discussed: 1. structure and safety analysis, 2. aging phenomena, 3. nondestructive testing, and 4. optimization of in-service inspection

  17. Finite test sets development method for test execution of safety critical software

    International Nuclear Information System (INIS)

    El-Bordany Ayman; Yun, Won Young

    2014-01-01

    It reads inputs, computes new states, and updates output for each scan cycle. Korea Nuclear Instrumentation and Control System (KNICS) has recently developed a fully digitalized Reactor Protection System (RPS) based on PLD. As a digital system, this RPS is equipped with a dedicated software. The Reliability of this software is crucial to NPPs safety where its malfunction may cause irreversible consequences and affect the whole system as a Common Cause Failure (CCF). To guarantee the reliability of the whole system, the reliability of this software needs to be quantified. There are three representative methods for software reliability quantification, namely the Verification and Validation (V and V) quality-based method, the Software Reliability Growth Model (SRGM), and the test-based method. An important concept of the guidance is that the test sets represent 'trajectories' (a series of successive values for the input variables of a program that occur during the operation of the software over time) in the space of inputs to the software.. Actually, the inputs to the software depends on the state of plant at that time, and these inputs form a new internal state of the software by changing values of some variables. In other words, internal state of the software at specific timing depends on the history of past inputs. Here the internal state of the software which can be changed by past inputs is named as Context of Software (CoS). In a certain CoS, a software failure occurs when a fault is triggered by some inputs. To cover the failure occurrence mechanism of a software, preceding researches insist that the inputs should be a trajectory form. However, in this approach, there are two critical problems. One is the length of the trajectory input. Input trajectory should long enough to cover failure mechanism, but the enough length is not clear. What is worse, to cover some accident scenario, one set of input should represent dozen hours of successive values

  18. The out-of-pile test for internal pressure measurement of nuclear fuel rod using LVDT

    Energy Technology Data Exchange (ETDEWEB)

    Son, J. M.; Kim, B. K.; Kim, D. S.; Joo, K. N.; Park, S. J.; Kang, Y. H.; Kim, Y. K.; Yeum, K. I. [KAERI, Taejon (Korea, Republic of)

    2002-05-01

    As a part of the development of instrumentation technologies for the nuclear fuel irradiation test in HANARO(High-flux Advanced Nuclear Application Reactor), the internal pressure measurement technique of the nuclear fuel rod is being developed using LVDT(Linear Variable Differential Transformer). The objectives of this test were to understand the LVDT's characteristics and to study its application techniques for fuel irradiation technology. It will be required to analyze the acquired internal pressure of fuel rod during fuel irradiation test in HANARO. Therefore, the out of pile test system for pressure measurement was developed, and the test with the LVDT at room temperature were performed. This test were implemented in 1 kg/cm{sup 2} increment from 1 kg/cm{sup 2} to 30 kg/cm{sup 2}, and repeated 6 times at same condition. The LVDT's sensitivities were obtained by following two ways, the one by test and the other by calculation from characteristics data. These two sensitivities were compared and analyzed. The calculation method for internal pressure of nuclear fuel rod at specified temperature was also established. The results of the out-of-pile test will be used to predict accurately the internal pressure of fuel rod during irradiation test. And, the well qualified out-of-pile tests are needed to understand the LVDT's detail characteristics at high temperature for the detail design of the fuel irradiation capsule.

  19. 77 FR 75699 - Pipeline Safety: Reporting of Exceedances of Maximum Allowable Operating Pressure

    Science.gov (United States)

    2012-12-21

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket No... AGENCY: Pipeline and Hazardous Materials Safety Administration (PHMSA); DOT. ACTION: Notice; Issuance of... occurs. This reporting requirement is applicable to all gas transmission pipeline facility owners and...

  20. SPES-2, AP600 intergral system test S01007 2 inch CL to core make-up tank pressure balance line break

    Energy Technology Data Exchange (ETDEWEB)

    Bacchiani, M.; Medich, C.; Rigamonti, M. [SIET S.p.A. Piacenza (Italy)] [and others

    1995-09-01

    The SPES-2 is a full height, full pressure experimental test facility reproducing the Westinghouse AP600 reactor with a scaling factor of 1/395. The experimental plant, designed and operated by SIET in Piacenza, consists of a full simulation of the AP600 primary core cooling system including all the passive and active safety systems. In 1992, Westinghouse, in cooperation with ENEL (Ente Nazionale per l` Energia Elettrica), ENEA (Enter per le numove Technlogie, l` Energia e l` Ambient), Siet (Societa Informazioni Esperienze Termoidraulich) and ANSALDO developed an experimental program to test the integrated behaviour of the AP600 passive safety systems. The SPES-2 test matrix, concluded in November 1994, has examined the AP600 passive safety system response for a range of small break LOCAs at different locations on the primary system and on the passive system lines; single steam generator tube ruptures with passive and active safety systems and a main steam line break transient to demonstrate the boration capability of passive safety systems for rapid cooldown. Each of the tests has provided detailed experimental results for verification of the capability of the analysis methods to predict the integrated passive safety system behaviour. Cold and hot shakedown tests have been performed on the facility to check the characteristics of the plant before starting the experimental campaign. The paper first presents a description of the SPES-2 test facility then the main results of S01007 test {open_quotes}2{close_quotes} Cold Leg (CL) to Core Make-up Tank (CMT) pressure balance line break{close_quotes} are reported and compared with predictions performed using RELAP5/mod3/80 obtained by ANSALDO through agreement with U.S.N.R.C. (U.S. Nuclear Regulatory Commission). The SPES-2 nodalization and all the calculations here presented were performed by ANSALDO and sponsored by ENEL as a part of pre-test predictions for SPES-2.