WorldWideScience

Sample records for surrounding pressure vessel

  1. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

    Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

  2. PRESSURE-RESISTANT VESSEL

    NARCIS (Netherlands)

    Beukers, A.; De Jong, T.

    1997-01-01

    Abstract of WO 9717570 (A1) The invention is directed to a wheel-shaped pressure-resistant vessel for gaseous, liquid or liquefied material having a substantially rigid shape, said vessel comprising a substantially continuous shell of a fiber-reinforced resin having a central opening, an inner l

  3. Pressurized Vessel Slurry Pumping

    Energy Technology Data Exchange (ETDEWEB)

    Pound, C.R.

    2001-09-17

    This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air.

  4. GOLD PRESSURE VESSEL SEAL

    Science.gov (United States)

    Smith, A.E.

    1963-11-26

    An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

  5. 46 CFR 169.249 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels. 169.249 Section 169.249 Shipping COAST... and Certification Inspections § 169.249 Pressure vessels. Pressure vessels must meet the requirements of part 54 of this chapter. The inspection procedures for pressure vessels are contained in...

  6. 46 CFR 182.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels. 182.330 Section 182.330 Shipping COAST...) MACHINERY INSTALLATION Auxiliary Machinery § 182.330 Pressure vessels. All unfired pressure vessels must be... unfired pressure vessels must meet the applicable requirements of subchapter F (Marine Engineering)...

  7. Pressure vessel and method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Saunders, Timothy

    2017-09-05

    A pressure vessel includes a pump having a passage that extends between an inlet and an outlet. A duct at the pump outlet includes at least one dimension that is adjustable to facilitate forming a dynamic seal that limits backflow of gas through the passage.

  8. 46 CFR 119.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels. 119.330 Section 119.330 Shipping COAST... Machinery § 119.330 Pressure vessels. All unfired pressure vessels must be installed to the satisfaction of the cognizant OCMI. The design, construction, and original testing of such unfired pressure...

  9. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and pressure piping. 197.462 Section... Diving Equipment § 197.462 Pressure vessels and pressure piping. (a) The diving supervisor shall ensure that each pressure vessel, including each volume tank, cylinder and PVHO, and each pressure...

  10. Unilateral lung agenesis--detrimental roles of surrounding vessels.

    Science.gov (United States)

    Chou, An-Kou; Huang, Shu-Chien; Chen, Shyh-Jye; Huang, Pei-Ming; Wang, Jou-Kou; Wu, Mei-Hwan; Chen, Yih-Sharng; Chang, Chung-I; Chiu, Ing-Sh; Wu, En-Ting

    2007-03-01

    Unilateral lung agenesis is a rare congenital defect and could be associated with multiple abnormalities. The patients usually have poor long-term outcomes especially in those with right lung agenesis. We reviewed the 10-year experience in our hospital to describe special clinical features and try to delineate the causes of poor outcomes. From 1995 to 2005, 14 patients less than 18 years of age with unilateral lung agenesis (4 with left agenesis, 10 with right agenesis) were enrolled. Medical records reviewed included diagnosis, presentation, chromosome anomalies, cardiovascular anomalies and interventions, outcomes. We found that the mechanisms of severe airway disease in right lung agenesis included (1) trachea compression by the aortic arch, (2) the presence of "pseudo-ring-sling complex," (3) distended pulmonary artery due to left to right shunt which impinged the only bronchus, and finally (4) the persistent LSVC that restricts the growth of trachea. The etiologies of airway complication in left lung agenesis included anomalous aortic arch compression on trachea and the coexisting heart disease with significant left to right shunt, which impinged on the bronchus. In conclusion, unilateral lung agenesis has frequently associated airway problems due to its surrounding vessels. Satisfactory airway intervention remains challenging. This disease still requires great effort to improve patient outcomes.

  11. Hydrogen storage in insulated pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

    1998-08-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

  12. Organic fiber/epoxy pressure vessels

    Science.gov (United States)

    Chiao, T. T.; Hamstad, M. A.; Marcon, M. A.

    1974-01-01

    We evaluated the performance of an organic fiber in an epoxy matrix by winding 20-cm diam spherical and cylindrical pressure vessels of various designs. For the spherical vessels, we used soft aluminum liners 0.76 mm thick for a double boss design and 2 mm thick for a single boss design. For the cylindrical vessels, we used both 0.5-mm rubber liners and 0.76-mm soft aluminum liners. Vessels of both types were tested for burst pressure and cyclic fatigue at room temperature and liquid hydrogen temperature. The effects of temperature and vessel shape on the vessel performance factor were negligible. Our vessel fatigue data were marred by premature failure of the liners.

  13. High Toughness Light Weight Pressure Vessel Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Proposed is a pressure vessel with 25% better Fracture Strength over equal strength designed Fiberglass to help reduce 10 to 25% weight for aerospace use. Phase I is...

  14. Liquid Nitrogen Subcooler Pressure Vessel Engineering Note

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, R.; /Fermilab

    1997-04-24

    The normal operating pressure of this dewar is expected to be less than 15 psig. This vessel is open to atmospheric pressure thru a non-isolatable vent line. The backpressure in the vent line was calculated to be less than 1.5 psig at maximum anticipated flow rates.

  15. Blood vessels, circulation and blood pressure.

    Science.gov (United States)

    Hendry, Charles; Farley, Alistair; McLafferty, Ella

    This article, which forms part of the life sciences series, describes the vessels of the body's blood and lymphatic circulatory systems. Blood pressure and its regulatory systems are examined. The causes and management of hypertension are also explored. It is important that nurses and other healthcare professionals understand the various mechanisms involved in the regulation of blood pressure to prevent high blood pressure or ameliorate its damaging consequences.

  16. 46 CFR 50.30-20 - Class III pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class III pressure vessels. 50.30-20 Section 50.30-20... Fabrication Inspection § 50.30-20 Class III pressure vessels. (a) Class III pressure vessels shall be subject... specifically exempted by other regulations in this subchapter. (b) For Class III welded pressure vessels,...

  17. 46 CFR 50.30-15 - Class II pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class II pressure vessels. 50.30-15 Section 50.30-15... Fabrication Inspection § 50.30-15 Class II pressure vessels. (a) Class II pressure vessels shall be subject to... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the...

  18. 46 CFR 61.10-5 - Pressure vessels in service.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessels in service. 61.10-5 Section 61.10-5... INSPECTIONS Tests and Inspections of Pressure Vessels § 61.10-5 Pressure vessels in service. (a) Basic requirements. Each pressure vessel must be examined or tested every 5 years. The extent of the test...

  19. Characterizing Acoustic Sources in Pressure Vessels

    Institute of Scientific and Technical Information of China (English)

    李路明; 郑鹏; 刘时风; 施克仁

    2002-01-01

    The "dream" of acoustic emission (AE) testing is to get the acoustic source characteristics from AE signals, especially when evaluating aging pressure vessels. In this paper, the wavelet transform was used to analyze different AE signals from cracks (surface and inner), pencil-lead-breakage and leakage. These acoustic sources were applied on an actual pressure vessel. While the vessel experienced hydraulic pressure, their AE signals were acquired by a digital AE testing system with a wide frequency band transducer and a high speed A/D converter. Then, the digital signals were analyzed using the wavelet transform method. Correlation coefficients of the transformed data show that the different acoustic sources can be easily identified.

  20. Pressure vessel inspections using ultrasonic phased arrays

    Energy Technology Data Exchange (ETDEWEB)

    Moles, M.D.C. [R/D Tech Toronto, Toronto, ON (Canada); Jacques, F.; Dube, N. [R/D Tech Quebec PQ (Canada)

    2003-07-01

    Pressure vessels are used in several industries, including the petrochemical and petroleum industries. Welds in the pressure vessels often produce defects which propagate and fail with time. Many types of pressure vessel weld inspections can now be conducted using automated ultrasonics which offers several advantages over radiography. In terms of phased arrays, custom tailored time-of-flight diffraction (TOFD) and ASME code Section V raster scans, can successfully perform high speed inspections with minimal operator subjectivity. The phased array beams can be steered, scanned, swept and focused electronically. Beam steering can be used to map welds at appropriate angles to optimize probability of defect detection. The main challenge lies with the initial equipment cost, technology awareness, and availability of trained operators. Phased arrays offer great flexibility for different components, defect detection and tailored imaging. 8 refs., 10 figs.

  1. Curved and conformal high-pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Croteau, Paul F.; Kuczek, Andrzej E.; Zhao, Wenping

    2016-10-25

    A high-pressure vessel is provided. The high-pressure vessel may comprise a first chamber defined at least partially by a first wall, and a second chamber defined at least partially by the first wall. The first chamber and the second chamber may form a curved contour of the high-pressure vessel. A modular tank assembly is also provided, and may comprise a first mid tube having a convex geometry. The first mid tube may be defined by a first inner wall, a curved wall extending from the first inner wall, and a second inner wall extending from the curved wall. The first inner wall may be disposed at an angle relative to the second inner wall. The first mid tube may further be defined by a short curved wall opposite the curved wall and extending from the second inner wall to the first inner wall.

  2. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  3. Cavitation distribution within large phantom vessel and mechanical damage formed on surrounding vessel wall.

    Science.gov (United States)

    Qiao, Yangzi; Yin, Hui; Li, Zhaopeng; Wan, Mingxi

    2013-11-01

    Blood vessel is one of the most important targets encountered during focused ultrasound (FU) therapy. The lasting high temperature caused by continuous FU can result in structural modification of small vessel. For the vessel with a diameter larger than 2mm, convective cooling can significantly weaken the thermal effect of FU. Meanwhile, the continued presence of ultrasound will cause repetitive cavitation and acoustic microstreaming, making comprehension of continuous wave induced cavitation effect in large vessels necessary. The Sonoluminescence (SL) method, mechanical damage observation and high-speed camera were used in this study to investigate the combination effect of ultrasound contrast agents (UCAs) and continuous FU in large phantom vessels with a diameter of 10mm without consideration of thermal effect. When the focus was positioned at the proximal wall, cylindrical hole along the acoustic axis opposite the ultrasound wave propagation direction was observed at the input power equal to or greater than 50 W. When the focus was located at the distal wall, only small tunnels can be found. The place where the cylindrical hole formed was corresponding to where bubbles gathered and emitted brilliant light near the wall. Without UCAs neither such bright SL nor cylindrical hole can be found. However, the UCAs concentration had little influence on the SL distribution and the length of cylindrical hole. The SL intensity near the proximal vessel wall and the length of the cylindrical hole both increased with the input power. It is suggested that these findings need to be considered in the large vessel therapy and UCAs usage.

  4. Simulating the Mineral Scale by High Pressure Thermal Vessel

    Science.gov (United States)

    Huang, Y. H.; Liu, H. L.; Chen, H. F.; Song, S. R.

    2014-12-01

    The generating capacity of Chingshui geothermal power plant decreased rapidly after it had operated three years. Chinese Petroleum Corporation (CPC) attributed the main reason was the depletion of reservoir. One reason was that the reservoir did not be recharged. And the other was the mineral scale in reservoir and pipes which caused flow rate decreased. There are abundant geothermal energy in Taiwan. But in Chingshui, the spring has amount content of carbonate. Most scaling are calcium carbonate and silica. These two materials have different solubility in various pH and physical conditions. Because the pressure reduced in the process of upwelling, the hot spring from the reservoir deposited calcium carbonate immediately by large carbon dioxide escape. This result caused the diameter of pipeline reduced. Besides, as the temperature decreased, the silica would scaling in the part of heat exchanger. To avoid the failure experience in Chingshui , how to prevent the mineral scaling is the key point that we need to solve. Our study will use hydrothermal experiments by High Pressure Thermal Vessel to simulate the process of spring water upwelling from reservoir to surface, to understand whether calcium carbonate and silica scaling or not in different temperature and pressure. This study choose the Hongchailin well as objects to simulate, and the target layers of drilling well were set as Szeleng sandstone and Lushan slate. We used pure water and saturated water pressure in our experiments. There were four vessels in High thermal vessel. The first vessel was used to simulate the condition of reservoir. The second and third vessel were simulated the conditions in the well when spring water upwelling to the surface. And the last vessel was simulated the conditions on surface surroundings. We hope to get the temperature and pressure when the scaling occurred, and verified with the computing result, thus we can inhibit the scaling.

  5. Kendall Analysis of Cannon Pressure Vessels

    Science.gov (United States)

    2012-04-11

    To) 4. TITLE AND SUBTITLE New PVD Technologies for New Ordnance Coatings 5a. CONTRACT NUMBER W911NF-11-D-0001 5b. GRANT NUMBER...yield pressure; autofrettage; fatigue life; cannon pressure vessels; residual stress; Bauschinger effect; 16. SECURITY CLASSIFICATION OF: 17...documents, enter the title classification in parentheses. 5a. CONTRACT NUMBER. Enter all contract numbers as they appear in the report, e.g. F33615-86

  6. 46 CFR 115.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL PASSENGER VESSELS CARRYING MORE... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be...

  7. (Irradiation embrittlement of reactor pressure vessels)

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  8. Conformable pressure vessel for high pressure gas storage

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

    2016-01-12

    A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

  9. Estimation of ex-vessel steam explosion pressure loads

    Energy Technology Data Exchange (ETDEWEB)

    Leskovar, Matjaz [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)], E-mail: matjaz.leskovar@ijs.si; Ursic, Mitja [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2009-11-15

    An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel-coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.

  10. Compact insert design for cryogenic pressure vessels

    Science.gov (United States)

    Aceves, Salvador M.; Ledesma-Orozco, Elias Rigoberto; Espinosa-Loza, Francisco; Petitpas, Guillaume; Switzer, Vernon A.

    2017-06-14

    A pressure vessel apparatus for cryogenic capable storage of hydrogen or other cryogenic gases at high pressure includes an insert with a parallel inlet duct, a perpendicular inlet duct connected to the parallel inlet. The perpendicular inlet duct and the parallel inlet duct connect the interior cavity with the external components. The insert also includes a parallel outlet duct and a perpendicular outlet duct connected to the parallel outlet duct. The perpendicular outlet duct and the parallel outlet duct connect the interior cavity with the external components.

  11. 46 CFR 58.60-3 - Pressure vessel.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessel. 58.60-3 Section 58.60-3 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING MAIN AND AUXILIARY MACHINERY AND... Pressure vessel. A pressure vessel that is a component in an industrial system under this subpart must...

  12. Computing the partial volume of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Wiencke, Bent [Nestle USA, Corporate Engineering, 800 N. Brand Blvd, Glendale, CA 91203 (United States)

    2010-06-15

    The computation of the partial and total volume of pressure vessels with various type of head profiles requires detailed knowledge of the head profile geometry. Depending on the type of head profile the derivation of the equations can become very complex and the calculation process cumbersome. Certain head profiles require numerical methods to obtain the partial volume, which for most application is beyond the scope of practicability. This paper suggests a unique method that simplifies the calculation procedure for the various types of head profiles by using one common set of equations without the need for numerical or complex computation methods. For ease of use, all equations presented in this paper are summarized in a single table format for horizontal and vertical vessels. (author)

  13. 33 CFR 165.1411 - Security zone; waters surrounding U.S. Forces vessel SBX-1, HI.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Security zone; waters surrounding U.S. Forces vessel SBX-1, HI. 165.1411 Section 165.1411 Navigation and Navigable Waters COAST GUARD... § 165.1411 Security zone; waters surrounding U.S. Forces vessel SBX-1, HI. (a) Location. The...

  14. Structural integrity of nuclear reactor pressure vessels

    Science.gov (United States)

    Knott, John F.

    2013-09-01

    The paper starts from concerns expressed by Sir Alan Cottrell, in the early 1970s, related to the safety of the pressurized water reactor (PWR) proposed at that time for the next phase of electrical power generation. It proceeds to describe the design and operation of nuclear generation plant and gives details of the manufacture of PWR reactor pressure vessels (RPVs). Attention is paid to stress-relief cracking and under-clad cracking, experienced with early RPVs, explaining the mechanisms for these forms of cracking and the means taken to avoid them. Particular note is made of the contribution of non-destructive inspection to structural integrity. Factors affecting brittle fracture in RPV steels are described: in particular, effects of neutron irradiation. The use of fracture mechanics to assess defect tolerance is explained, together with the failure assessment diagram embodied in the R6 procedure. There is discussion of the Master Curve and how it incorporates effects of irradiation on fracture toughness. Dangers associated with extrapolation of data to low probabilities are illustrated. The treatment of fatigue-crack growth is described, in the context of transients that may be experienced in the operation of PWR plant. Detailed attention is paid to the thermal shock associated with a large loss-of-coolant accident. The final section reviews the arguments advanced to justify 'Incredibility of Failure' and how these are incorporated in assessments of the integrity of existing plant and proposed 'new build' PWR pressure vessels.

  15. Pressure vessel calculations for VVER-440 reactors.

    Science.gov (United States)

    Hordósy, G; Hegyi, Gy; Keresztúri, A; Maráczy, Cs; Temesvári, E; Vértes, P; Zsolnay, E

    2005-01-01

    For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.

  16. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Laboratory

    2008-04-01

    The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

  17. Midland reactor pressure vessel flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Foulds, J.R.; Kennedy, E.L. [Failure Analysis Associates, Inc., Menlo Park, CA (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States)

    1993-12-01

    The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center`s (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions.

  18. Reactor pressure vessel structural integrity research

    Energy Technology Data Exchange (ETDEWEB)

    Pennell, W.E.; Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  19. Treating asphericity in fuel particle pressure vessel modeling

    Science.gov (United States)

    Miller, Gregory K.; Wadsworth, Derek C.

    1994-07-01

    The prototypical nuclear fuel of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR) consists of spherical TRISO-coated particles suspended in graphite cylinders. The coating layers surrounding the fuel kernels in these particles consist of pyrolytic carbon layers and a silicon carbide layer. These coating layers act as a pressure vessel which retains fission product gases. In the operating conditions of the NP-MHTGR, a small percentage of these particles (pressure vessels) are expected to fail due to the pressure loading. The fuel particles of the NP-MHTGR deviate to some degree from a true spherical shape, which may have some effect on the failure percentages. A method is presented that treats the asphericity of the particles in predicting failure probabilities for particle samples. It utilizes a combination of finite element analysis and Monte Carlo sampling and is based on the Weibull statistical theory. The method is used here to assess the effects of asphericity in particles of two common geometric shapes, i.e. faceted particles and ellipsoidal particles. The method presented could be used to treat particle anomalies other than asphericity.

  20. Design and implementation of a visual monitoring system to ensure safety in the water surrounding a container vessel

    Institute of Scientific and Technical Information of China (English)

    JIN Yong-xing; WANG Ze-sheng; CHEN Jin-biao; BUPing

    2008-01-01

    Container vessels navigate among the world's ports, frequently passing through narrow and congested waters. Due to the many layers of containers on a container vessel's decks, it is difficult for the crew to be aware of all fishing vessels and other obstacles in a container vessel's radar observation blind zone. This greatly increases the risk of collisions and other accidents. Given such great challenges to safe navigation and safety management with container vessels, their security risks are severe. An effective visual monitoring system can improve the safety of the water area surrounding container vessel by eliminating a vessel's observation blind zone, providing an effective safety measure for vessels navigating fishing zones and other troublesome areas. The system has other functions, such as accident recording, ship security, and monitoring of loading and unloading operations, thus ensuring the ship operates safely. Six months' trial operation showed that the system facilitates safe navigation of container vessels.

  1. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  2. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10-2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006-2007 in a geometry representative of 1300 MWe PWR.

  3. NDE and Stress Monitoring on Composite Overwrapped Pressure Vessels Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Damage caused by composite overwrapped pressure vessels (COPVs) failure can be catastrophic. Thus, monitoring condition and stress in the composite overwrap,...

  4. Radiation effects on reactor pressure vessel supports

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, R.E. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Lipinski, R.E. [Idaho National Engineering Lab., Rockville, MD (United States)

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue.

  5. 46 CFR 176.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a.... (b) Periodic inspection and testing requirements for boilers are contained in § 61.05 in subchapter...

  6. Explosion pressures of hydrocarbon-air mixtures in closed vessels.

    Science.gov (United States)

    Razus, Domnina; Movileanu, Codina; Brinzea, Venera; Oancea, D

    2006-07-31

    An experimental study on pressure evolution during closed vessel explosions of several gaseous fuel-air mixtures was performed, at various initial pressures within 0.3-1.2 bar and ambient initial temperature. Explosion pressures and explosion times are reported for methane-, n-pentane-, n-hexane-, propene-, butene-, butadiene-, cyclohexane- and benzene-air mixtures. The explosion pressures measured in a spherical vessel (Phi=10 cm) and in three cylindrical vessels with different diameter/height ratios are examined in comparison with the adiabatic explosion pressures, computed by assuming chemical equilibrium within the flame front. The influence of initial pressure, fuel concentration and heat losses during propagation (determined by the size and shape of the explosion vessel and by the position of the ignition source) on explosion pressures and explosion times are discussed for some of the examined systems.

  7. Fatigue life of organic fiber/epoxy pressure vessels

    Science.gov (United States)

    Hamstad, M. A.; Chiao, T. T.; Patterson, R. G.

    1975-01-01

    The cyclic fatigue life of 10.2-cm-diam cylindrical pressure vessels has been studied. The vessels were made of an organic fiber/epoxy composite. To determine the typical strength distribution of the vessels, 25 of them were internally pressurized until they burst. Twenty-five vessels were then tested under sinusoidal cycling at 1 Hz between 4% and 91% of the mean burst strength. An additional twenty-five vessels were tested between 4% and 91% with a rectangular pressure pulse at 1/3 Hz. A limited number of vessels were tested for stress rupture at the 91% level. Cyclic life was found to depend on time at peak load as well as the number of stress cycles.

  8. Evaluation of insulated pressure vessels for cryogenic hydrogen storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Garcia-Villazana, O; Martinez-Frias, J

    1999-03-01

    This paper presents an analytical and experimental evaluation of the applicability of insulated pressure vessels for hydrogen-fueled light-duty vehicles. Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH?) or ambient-temperature compressed hydrogen (CH2). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). The purpose of this work is to verify that commercially available aluminum-lined, fiber- wrapped vessels can be used for cryogenic hydrogen storage. The paper reports on previous and ongoing tests and analyses that have the purpose of improving the system design and assure its safety.

  9. Firefighter's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

    The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

  10. ASME code ductile failure criteria for impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, Robert E.; Duffey, T. A. (Thomas A.); Rodriguez, E. A. (Edward A.)

    2003-01-01

    Ductile failure criteria suitable for application to impulsively loaded high pressure vessels that are designed to the rules of the ASME Code Section VI11 Division 3 are described and justified. The criteria are based upon prevention of load instability and the associated global failure mechanisms, and on protection against progressive distortion for multiple-use vessels. The criteria are demonstrated by the design and analysis of vessels that contain high explosive charges.

  11. Quantification of Processing Effects on Filament Wound Pressure Vessels

    Science.gov (United States)

    Aiello, Robert A.; Chamis, Christos C.

    1999-01-01

    A computational simulation procedure is described which is designed specifically for the modeling and analysis of filament wound pressure vessels. Cylindrical vessels with spherical or elliptical end caps can be generated automatically. End caps other than spherical or elliptical may be modeled by varying circular sections along the x-axis according to the C C! end cap shape. The finite element model generated is composed of plate type quadrilateral shell elements on the entire vessel surface. This computational procedure can also be sued to generate grid, connectivity and material cards (bulk data) for component parts of a larger model. These bulk data are assigned to a user designated file for finite element structural/stress analysis of composite pressure vessels. The procedure accommodates filament would pressure vessels of all types of shells-of-revolution. It has provisions to readily evaluate initial stresses due to pretension in the winding filaments and residual stresses due to cure temperature.

  12. Quantification of Processing Effects on Filament Wound Pressure Vessels. Revision

    Science.gov (United States)

    Aiello, Robert A.; Chamis, Christos C.

    2002-01-01

    A computational simulation procedure is described which is designed specifically for the modeling and analysis of filament wound pressure vessels. Cylindrical vessels with spherical or elliptical end caps can be generated automatically. End caps other than spherical or elliptical may be modeled by varying circular sections along the x-axis according to the end cap shape. The finite element model generated is composed of plate type quadrilateral shell elements on the entire vessel surface. This computational procedure can also be used to generate grid, connectivity and material cards (bulk data) for component parts of a larger model. These bulk data are assigned to a user designated file for finite element structural/stress analysis of composite pressure vessels. The procedure accommodates filament wound pressure vessels of all types of shells-of -revolution. It has provisions to readily evaluate initial stresses due to pretension in the winding filaments and residual stresses due to cure temperature.

  13. Evaluation of pressure vessel fluence computation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. H.; Whang, I. S.; Kim, T. G.; Lee, H. C. [Seoul National Univ., Seoul (Korea, Republic of); Jang, M. H.; Whang, H. R.; Park, W. S.; An, J. G. [Korea Atomic Energy Research Insitute, Daejeon (Korea, Republic of)

    1994-04-15

    This study was performed as follows: evaluation of neutron fluence calculational methodology through the analysis of benchmark problem, evaluation of calculational results of Yonggwang 3 and 4 reactor vessel fluence, examination of calculational results against the requirements by 10CFR 50.61 and/or standard review plan. The preservation of reactor vessel integrity throughout the reactor lifetime is directly related to the economical and safe operation of nuclear power plants. In this regard, it is very important to accurately predict and assess the neutron fluence which impacts directly upon the reactor vessel integrity. The accurate. prediction and assessment of the reactor vessel fluence require the use of accurate data as well as systematic methodology. However, it is felt that all of these prerequisites are not sufficient at the moment. It is, therefore, recommended to establish a systematic methodology with sufficient nuclear data library for the reliable licensing review of the reactor vessel safety, by performing R and D to resolve the problems presented in this study and by using the results of this study.

  14. Interpretation of Strain Measurements on Nuclear Pressure Vessels

    DEFF Research Database (Denmark)

    Andersen, Svend Ib Smidt; Engbæk, Preben

    1980-01-01

    Selected results from strain measurements on four nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts. The resu......Selected results from strain measurements on four nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts....... The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as a detailed knowledge of the behaviour of the signal...... with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem...

  15. Nuclear reactor pressure vessel-specific flaw distribution development

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.

    1992-09-01

    Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses.

  16. Nuclear reactor pressure vessel-specific flaw distribution development

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.

    1992-01-01

    Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses.

  17. Influence of residual stresses on failure pressure of cylindrical pressure vessels

    Institute of Scientific and Technical Information of China (English)

    M. Jeyakumar; T. Christopher

    2013-01-01

    The utilization of pressure vessels in aerospace applications is manifold. In this work, finite element analysis (FEA) has been carried out using ANSYS software package with 2D axisym-metric model to access the failure pressure of cylindrical pressure vessel made of ASTM A36 carbon steel having weld-induced residual stresses. To find out the effect of residual stresses on failure pressure, first an elasto-plastic analysis is performed to find out the failure pressure of pressure vessel not having residual stresses. Then a thermo-mechanical finite element analysis is performed to assess the residual stresses developed in the pressure vessel during welding. Finally one more elasto-plastic analysis is performed to assess the effect of residual stresses on failure pressure of the pressure vessel having residual stresses. This analysis indicates reduction in the failure pressure due to unfavorable residual stresses.

  18. Calculations of plastic collapse load of pressure vessel using FEA

    Institute of Scientific and Technical Information of China (English)

    Peng-fei LIU; Jin-yang ZHENG; Li MA; Cun-jian MIAO; Lin-lin WU

    2008-01-01

    This paper proposes a theoretical method using finite element analysis (FEA) to calculate the plastic collapse loads of pressure vessels under internal pressure, and compares the analytical methods according to three criteria stated in the ASME Boiler Pressure Vessel Code. First, a finite element technique using the arc-length algorithm and the restart analysis is developed to conduct the plastic collapse analysis of vessels, which includes the material and geometry non-linear properties of vessels. Second,as the mechanical properties of vessels are assumed to be elastic-perfectly plastic, the limit load analysis is performed by employing the Newton-Raphson algorithm, while the limit pressure of vessels is obtained by the twice-elastic-slope method and the tangent intersection method respectively to avoid excessive deformation. Finally, the elastic stress analysis under working pressure is conducted and the stress strength of vessels is checked by sorting the stress results. The results are compared with those obtained by experiments and other existing models. This work provides a reference for the selection of the failure criteria and the calculation of the plastic collapse load.

  19. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

    Taarnhøj, Nina C B B; Larsen, Michael; Sander, Birgit

    2006-01-01

    PURPOSE: To assess the relative influence of genetic and environmental effects on retinal vessel diameters and blood pressure in healthy adults, as well as the possible genetic connection between these two characteristics. METHODS: In 55 monozygotic and 50 dizygotic same-sex healthy twin pairs......%-80%) for CRAE, 83% (95% CI: 73%-89%) for CRVE, and 61% (95% CI: 44%-73%) for mean arterial blood pressure (MABP). Retinal artery diameter decreased with increasing age and increasing arterial blood pressure. Mean vessel diameters in the population were 165.8 +/- 14.9 microm for CRAE, 246.2 +/- 17.7 microm...... and blood glucose, variations in retinal blood vessel diameters and blood pressure were predominantly attributable to genetic effects. A genetic influence may have a role in individual susceptibility to hypertension and other vascular diseases. The results suggest that retinal vessel diameters...

  20. Investigation of impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

    1963-10-15

    Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

  1. High pressure gas vessels for neutron scattering experiments

    CERN Document Server

    Done, R; Evans, B E; Bowden, Z A

    2010-01-01

    The combination of high pressure techniques with neutron scattering proves to be a powerful tool for studying the phase transitions and physical properties of solids in terms of inter-atomic distances. In our report we are going to review a high pressure technique based on a gas medium compression. This technique covers the pressure range up to ~0.7GPa (in special cases 1.4GPa) and typically uses compressed helium gas as the pressure medium. We are going to look briefly at scientific areas where high pressure gas vessels are intensively used in neutron scattering experiments. After that we are going to describe the current situation in high pressure gas technology; specifically looking at materials of construction, designs of seals and pressure vessels and the equipment used for generating high pressure gas.

  2. Structural analysis of in-pool pressure vessel in CNS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok Hoon; Lee, K. H.; Lee, J. H.; Lee, H. Y

    2005-08-15

    The in-pool pressure vessel in the Cold Neutron Source consists of the moderator vessel and the vacuum tube which is inserted in vertical hole located in the reflector region. It is necessary that the moderator vessel is designed to minimize its thickness within the extent to satisfy the stress intensity allowable limits not to shrink the cold neutron flux. The vacuum tube should be designed to endure the high pressure loads by the fracture of the moderator vessel and the overpressure of cover gas. The stress calculation was performed to verify the design of the moderator vessel and vacuum tube. The loads taken into account in this analysis are pressure during normal operation, seismic events and thermal expansion. The detail analyses for the moderator vessel and the vacuum tube will be carried out after deciding the loads through the thermal hydraulic analysis. For the detail analysis, the loads such as failure of the moderator vessel and overpressure in cover gas should be considered by the accident analysis. The calculated stresses satisfied the ASME SC-1 component design and analysis rules. In the buckling analysis, the structural integrity was also verified in the vacuum tube such a long cylinder type.

  3. Technical Appendix to Cryogenic Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Mulholland, G.T.; Rucinski, R.A; /Fermilab

    1990-02-22

    The 20,000 gls. Liquid Argon dewar stores up to 15,000 gls. of high purity (<1.0 ppm O{sub 2}, 0.999995) LAr for use in the Liquid Argon calorimeters of E740, the D0 collider detector, at elevation 707-feet. The dewar provides for the total detector volume of 11,000 gls and a 4,000 gls. storage inventory. The large gas volume ({ge}5,000 gls.) serves operational needs and guards against overfill concerns. The LAr dewar functions in two modes: (1) low pressure (16 psi relief) storage, and liquid and gas transfer operations to and from the low pressure (13 psi relief) detector cryostats, and (2) high pressure (65 psi relief) liquid transfer operations to and from a delivery trailer at elevation 743-feet. The storage function is intended to be long term and nonventing. The dewar is equipped with a 40 kW LN{sub 2} condenser that operates to maintain the pressure constant in the storage mode. This service exactly parallels the NeH{sub 2} and D{sub 2} storage dewar services provided at the 15-feet bubble chamber for its operation.

  4. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessel for human occupancy (PVHO). 54.01-17... PRESSURE VESSELS General Requirements § 54.01-17 Pressure vessel for human occupancy (PVHO). Pressure vessels for human occupancy (PVHO's) must meet the requirements of subpart B (Commercial Diving...

  5. Expert system for evaluating the safety of pressure vessels

    Institute of Scientific and Technical Information of China (English)

    Dong Zhibo; Lu Yafeng; Wei Yanhong; Yang Yongfu; Ma Rui; Guo Ping

    2009-01-01

    With more application of welding technology in important structures more attention was paid to the evaluation of the safety of welded structures, the life prediction and decision to repair the welded structures. Based on material fiacture mechanism and Chinese standard of safety evaluations of pressure vessels, an expert system was developed to evaluate the safety of welded pressure vessels. The system can analyze the weld defects in a pressure vessel, convert different kinds of defects into equivalent cracks and obtain their equivalent sizes. Furthermore, the system can calculate the stress and strain in the positions of weld defects and make decision on whether the defects are tolerable or not according to the code. When it is tolerable, the system will calculate the safety margin. The fatigue life can be predicted if the defects undergo fatigue load too. Moreover, data bases are built for storing mechanical properties of material and evaluated results.

  6. Selection of materials for pressure vessels and chemical plants

    Energy Technology Data Exchange (ETDEWEB)

    Huppertz, P.H.; Retter, A. (Linde A.G., Hoellriegelskreuth (Germany, F.R.). Werksgruppe Tieftemperatur und Verfahrenstechnik)

    1980-04-01

    The selection of materials for pressure vessels and chemical plants depends on a number of factors such as operating, operating temperature, operating medium, regulations in force in the country of the plant user concerned and manufacturing possibilities. The essay clearly explains how the above specified factors individually influence the selection of materials. The article also deals with the ranges of application of certain material groups such as unalloyed and low-alloy steels, fine-grained steels, austenitic chromium-nickel steels, unalloyed ferritic chromium steels and other materials. The article closes with remarks on the operational safety of pressure vessels.

  7. Industrial safety of pressure vessels - structural integrity point of view

    Directory of Open Access Journals (Sweden)

    Sedmak Aleksandar

    2016-01-01

    Full Text Available This paper presents different aspects of pressure vessel safety in the scope of industrial safety, focused to the chemical industry. Quality assurance, including application of PED97/23 has been analysed first, followed shortly by the risk assessment and in details by the structural integrity approach, which has been illustrated with three case studies. One important conclusion, following such an approach, is that so-called water proof testing can actually jeopardize integrity of a pressure vessel instead of proving it. [Projekat Ministarstva nauke Republike Srbije, br. TR 174004 i br. TR 33044

  8. SMART composite high pressure vessels with integrated optical fiber sensors

    Science.gov (United States)

    Blazejewski, Wojciech; Czulak, Andrzej; Gasior, Pawel; Kaleta, Jerzy; Mech, Rafal

    2010-04-01

    In this paper application of integrated Optical Fiber Sensors for strain state monitoring of composite high pressure vessels is presented. The composite tanks find broad application in areas such as: automotive industry, aeronautics, rescue services, etc. In automotive application they are mainly used for gaseous fuels storage (like CNG or compressed Hydrogen). In comparison with standard steel vessels, composite ones have many advantages (i.e. high mechanical strength, significant weight reduction, etc). In the present work a novel technique of vessel manufacturing, according to this construction, was applied. It is called braiding technique, and can be used as an alternative to the winding method. During braiding process, between GFRC layers, two types of optical fiber sensors were installed: point sensors in the form of FBGs as well as interferometric sensors with long measuring arms (SOFO®). Integrated optical fiber sensors create the nervous system of the pressure vessel and are used for its structural health monitoring. OFS register deformation areas and detect construction damages in their early stage (ensure a high safety level for users). Applied sensor system also ensured a possibility of strain state monitoring even during the vessel manufacturing process. However the main application of OFS based monitoring system is to detect defects in the composite structure. An idea of such a SMART vessel with integrated sensor system as well as an algorithm of defect detection was presented.

  9. Threaded insert for compact cryogenic-capable pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Loza, Francisco; Ross, Timothy O.; Switzer, Vernon A.; Aceves, Salvador M.; Killingsworth, Nicholas J.; Ledesma-Orozco, Elias

    2015-06-16

    An insert for a cryogenic capable pressure vessel for storage of hydrogen or other cryogenic gases at high pressure. The insert provides the interface between a tank and internal and external components of the tank system. The insert can be used with tanks with any or all combinations of cryogenic, high pressure, and highly diffusive fluids. The insert can be threaded into the neck of a tank with an inner liner. The threads withstand the majority of the stress when the fluid inside the tank that is under pressure.

  10. Threaded insert for compact cryogenic-capable pressure vessels

    Science.gov (United States)

    Espinosa-Loza, Francisco; Ross, Timothy O.; Switzer, Vernon A.; Aceves, Salvador M.; Killingsworth, Nicholas J.; Ledesma-Orozco, Elias

    2015-06-16

    An insert for a cryogenic capable pressure vessel for storage of hydrogen or other cryogenic gases at high pressure. The insert provides the interface between a tank and internal and external components of the tank system. The insert can be used with tanks with any or all combinations of cryogenic, high pressure, and highly diffusive fluids. The insert can be threaded into the neck of a tank with an inner liner. The threads withstand the majority of the stress when the fluid inside the tank that is under pressure.

  11. Compressibility measurements of gases using externally heated pressure vessels.

    Science.gov (United States)

    Presnall, D. C.

    1971-01-01

    Most of the data collected under conditions of high temperature and pressure have been determined using a thick-walled bomb of carefully measured and fixed volume which is externally heated by an electric furnace or a thermostatically controlled bath. There are numerous variations on the basic method depending on the pressure-temperature range of interest, and the particular gas or gas mixture being studied. The construction and calibration of the apparatus is discussed, giving attention to the pressure vessel, the volume of the bomb, the measurement of pressure, the control and measurement of temperature, and the measurement of the amount and composition of gas in the bomb.

  12. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

    Taarnhøj, Nina C B B; Larsen, Michael; Sander, Birgit

    2006-01-01

    PURPOSE: To assess the relative influence of genetic and environmental effects on retinal vessel diameters and blood pressure in healthy adults, as well as the possible genetic connection between these two characteristics. METHODS: In 55 monozygotic and 50 dizygotic same-sex healthy twin pairs......%-80%) for CRAE, 83% (95% CI: 73%-89%) for CRVE, and 61% (95% CI: 44%-73%) for mean arterial blood pressure (MABP). Retinal artery diameter decreased with increasing age and increasing arterial blood pressure. Mean vessel diameters in the population were 165.8 +/- 14.9 microm for CRAE, 246.2 +/- 17.7 microm...... for CRVE, and 0.67 +/- 0.05 microm for AVR. No significant influence on artery or vein diameters was found for gender, smoking, body mass index (BMI), total cholesterol, fasting blood glucose, or 2-hour oral glucose tolerance test values. CONCLUSIONS: In healthy young adults with normal blood pressure...

  13. Optimal design of pressure vessel using an improved genetic algorithm

    Institute of Scientific and Technical Information of China (English)

    Peng-fei LIU; Ping XU; Shu-xin HAN; Jin-yang ZHENG

    2008-01-01

    As the idea of simulated annealing(SA) is introduced into the fitness function,an improved genetic algorithm(GA) is proposed to perform the optimal design of a pressure vessel which aims to attain the minimum weight under burst pressure constraint.The actual burst pressure is calculated using the arc-length and restart analysis in finite element analysis(FEA).A penalty function in the fitness function is proposed to denl with the constrained problem.The erects of the population size and the number of generations in the GA on the weight and burst pressure of the vessel are explored.The optimization results using the proposed GA are also compared with those using the simple GA and the conventional Monte Carlo method.

  14. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  15. 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.

  16. USER SPECIFICATIONS FOR PRESSURE VESSELS AND TECHNICAL INTEGRITY

    Directory of Open Access Journals (Sweden)

    K.S. Johnston

    2012-01-01

    Full Text Available

    ENGLISH ABSTRACT: Specifications translated from user requirements are prescribed in an attempt to capture and incorporate best practices with regards to the design, fabrication, testing, and operation of pressure vessels. The question as to whether these requirements affect the technical integrity of pressure vessels is often a subjective matter. This paper examines typical user requirement specifications against technical integrity of pressure vessels.
    The paper draws on a survey of a convenience sample of practising engineers in a diversified petrochemical company. When compared with failures on selected pressure vessels recorded by Phillips and Warwick, the respondent feedback confirms the user specifications that have the highest impact on technical integrity.

    AFRIKAANSE OPSOMMING: Gebruikersbehoeftes word saamgevat in spesifikasies wat lei tot goeie praktyk vir ontwerp, vervaarding, toetsing en bedryf van drukvate. Subjektiwiteit van die gebruikersbehoeftes mag soms die tegniese integriteit van ‘n drukvat beinvloed.
    Die navorsing maak by wyse van monsterneming gebruik van die kennis van ingenieurs wat werk in ‘n gediversifiseerde petrochemiese bedryf. Die terugvoering bevestig dat bogenoemde spesifikasies inderdaad die grootste invloed het op tegniese integriteit.

  17. 46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 97.30-1 Section... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief...

  18. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 196.30-1... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the Chief...

  19. Circular cylinders and pressure vessels stress analysis and design

    CERN Document Server

    Vullo, Vincenzo

    2014-01-01

    This book provides comprehensive coverage of stress and strain analysis of circular cylinders and pressure vessels, one of the classic topics of machine design theory and methodology. Whereas other books offer only a partial treatment of the subject and frequently consider stress analysis solely in the elastic field, Circular Cylinders and Pressure Vessels broadens the design horizons, analyzing theoretically what happens at pressures that stress the material beyond its yield point and at thermal loads that give rise to creep. The consideration of both traditional and advanced topics ensures that the book will be of value for a broad spectrum of readers, including students in postgraduate, and doctoral programs and established researchers and design engineers. The relations provided will serve as a sound basis for the design of products that are safe, technologically sophisticated, and compliant with standards and codes and for the development of innovative applications.

  20. Reliability Considerations for Composite Overwrapped Pressure Vessels on Spacecraft

    Science.gov (United States)

    Murthy, Pappu L. N.; Gyekenyesi, John P.; Grimes-Ledesma, Lorie; Phoenix, S. L.

    2007-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are used to store gases under high pressure onboard spacecraft. These are used for a variety of purposes such as propelling liquid fuel etc, Kevlar, glass, Carbon and other more recent fibers have all been in use to overwrap the vessels. COPVs usually have a thin metallic liner with the primary purpose of containing the gases and prevent any leakage. The liner is overwrapped with filament wound composite such as Kevlar, Carbon or Glass fiber. Although the liner is required to perform in the leak before break mode making the failure a relatively benign mode, the overwrap can fail catastrophically under sustained load due to stress rupture. It is this failure mode that is of major concern as the stored energy of such vessels is often great enough ta cause loss of crew and vehicle. The present paper addresses some of the reliability concerns associated specifically with Kevlar Composite Overwrapped Pressure Vessels. The primary focus of the paper is on how reliability of COPV's are established for the purpose of deciding in general their flight worthiness and continued use. Analytical models based on existing design data will be presented showing how to achieve the required reliability metric to the end of a specific period of performance. Uncertainties in the design parameters and how they affect reliability and confidence intervals will be addressed as well. Some trade studies showing how reliability changes with time during a program period will be presented.

  1. Fabrication of toroidal composite pressure vessels. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Dodge, W.G.; Escalona, A.

    1996-11-24

    A method for fabricating composite pressure vessels having toroidal geometry was evaluated. Eight units were fabricated using fibrous graphite material wrapped over a thin-walled aluminum liner. The material was wrapped using a machine designed for wrapping, the graphite material was impregnated with an epoxy resin that was subsequently thermally cured. The units were fabricated using various winding patterns. They were hydrostatically tested to determine their performance. The method of fabrication was demonstrated. However, the improvement in performance to weight ratio over that obtainable by an all metal vessel probably does not justify the extra cost of fabrication.

  2. Improved Attachment in a Hybrid Inflatable Pressure Vessel

    Science.gov (United States)

    Johnson, Christopher J.; Patterson, Ross; Spexarth, Gary R.

    2010-01-01

    The vessel is a hybrid that comprises an inflatable shell attached to a rigid structure. The inflatable shell is, itself, a hybrid that comprises (1) a pressure bladder restrained against expansion by (2) a restraint layer that comprises a web of straps made from high-strength polymeric fabrics. The present improvements are intended to overcome deficiencies in those aspects of the original design that pertain to attachment of the inflatable shell to the rigid structure. In a typical intended application, such attachment(s) would be made at one or more window or hatch frames to incorporate the windows or hatches as integral parts of the overall vessel.

  3. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifing Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, Leigh

    2007-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle's Kevlar-49 fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed non-conservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic-plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23% lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  4. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifting Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, S. L.

    2009-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle s Kevlar-49 (DuPont) fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed nonconservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23 percent lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  5. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 7 2010-07-01 2010-07-01 false Portable air receivers and other unfired pressure vessels... § 1915.172 Portable air receivers and other unfired pressure vessels. (a) Portable, unfired pressure... vessels, or set to the lowest safe working pressure of the systems, whichever is lower. (d) A...

  6. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Report of repairs to boilers and pressure vessels. 109... Report of repairs to boilers and pressure vessels. Before making repairs, except normal repairs and maintenance such as replacement of valves or pressure seals, to boilers or unfired pressure vessels...

  7. ESTIMATE OF BURSTING PRESSURE OF MILD STEEL PRESSURE VESSEL AND PRESENTATION OF BURSTING FORMULA

    Institute of Scientific and Technical Information of China (English)

    ZHENG Chuanxiang

    2006-01-01

    In order to get more precise bursting pressure formula of mild steel, hundreds of bursting experiments of mild steel pressure vessels such as Q235(Gr.D) and 20R(1020) are done. Based on statistical data of bursting pressure and modification of Faupel formula, a more precise modified formula is given out according to the experimental data. It is proved to be more accurate after examining other bursting pressure value presented in many references. This bursting formula is very accurate in these experiments using pressure vessels with different diameter and shell thickness.Obviously, this modified bursting formula can be used in mild steel pressure vessels with different diameter and thickness of shell.

  8. Design Considerations For Blast Loads In Pressure Vessels.

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, E. A. (Edward A.); Nickell, Robert E.; Pepin, J. E. (Jason E.)

    2007-01-01

    Los Alamos National Laboratory (LANL), under the auspices of the U.S. Department of Energy (DOE) and the National Nuclear Security Administration (NNSA), conducts confined detonation experiments utilizing large, spherical, steel pressure vessels to contain the reaction products and hazardous materials from high-explosive (HE) events. Structural design and analysis considerations include: (a) Blast loading phase (i.e., impulsive loading); (b) Dynamic structural response; (c) Fragment (i.e., shrapnel) generation and penetration; (d) Ductile and non-ductile fracture; and (e) Design Criteria to ASME Code Sec. VIII, Div. 3, Impulsively Loaded Vessels. These vessels are designed for one-time-use only, efficiently utilizing the significant plastic energy absorption capability of ductile vessel materials. Alternatively, vessels may be designed for multiple-detonation events, in which case the material response is restricted to elastic or near-elastic range. Code of Federal Regulations, Title 10 Part 50 provides requirements for commercial nuclear reactor licensing; specifically dealing with accidental combustible gases in containment structures that might cause extreme loadings. The design philosophy contained herein may be applied to extreme loading events postulated to occur in nuclear reactor and non-nuclear systems or containments.

  9. Quality Testing of Gaseous Helium Pressure Vessels by Acoustic Emission

    CERN Document Server

    Barranco-Luque, M; Hervé, C; Margaroli, C; Sergo, V

    1998-01-01

    The resistance of pressure equipment is currently tested, before commissioning or at periodic maintenance, by means of normal pressure tests. Defects occurring inside materials during the execution of these tests or not seen by usual non-destructive techniques can remain as undetected potential sources of failure . The acoustic emission (AE) technique can detect and monitor the evolution of such failures. Industrial-size helium cryogenic systems employ cryogens often stored in gaseous form under pressure at ambient temperature. Standard initial and periodic pressure testing imposes operational constraints which other complementary testing methods, such as AE, could significantly alleviate. Recent reception testing of 250 m3 GHe storage vessels with a design pressure of 2.2 MPa for the LEP and LHC cryogenic systems has implemented AE with the above-mentioned aims.

  10. Stress Rupture Life Reliability Measures for Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Murthy, Pappu L. N.; Thesken, John C.; Phoenix, S. Leigh; Grimes-Ledesma, Lorie

    2007-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases onboard spacecraft. Kevlar (DuPont), glass, carbon and other more recent fibers have all been used as overwraps. Due to the fact that overwraps are subjected to sustained loads for an extended period during a mission, stress rupture failure is a major concern. It is therefore important to ascertain the reliability of these vessels by analysis, since the testing of each flight design cannot be completed on a practical time scale. The present paper examines specifically a Weibull statistics based stress rupture model and considers the various uncertainties associated with the model parameters. The paper also examines several reliability estimate measures that would be of use for the purpose of recertification and for qualifying flight worthiness of these vessels. Specifically, deterministic values for a point estimate, mean estimate and 90/95 percent confidence estimates of the reliability are all examined for a typical flight quality vessel under constant stress. The mean and the 90/95 percent confidence estimates are computed using Monte-Carlo simulation techniques by assuming distribution statistics of model parameters based also on simulation and on the available data, especially the sample sizes represented in the data. The data for the stress rupture model are obtained from the Lawrence Livermore National Laboratories (LLNL) stress rupture testing program, carried out for the past 35 years. Deterministic as well as probabilistic sensitivities are examined.

  11. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  12. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  13. EPRI activities to address reactor pressure vessel integrity issues

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.; Carter, R.G. [Electric Power Res. Inst., Charlotte, NC (United States)

    1999-12-01

    The demonstration of reactor pressure vessel (RPV) structural integrity is an essential element in ensuring the continued safe and reliable operation of US nuclear power plants. The Electric Power Research Institute (EPRI), through its domestic and international member utilities, continues to pursue an aggressive research program to develop technologies and capabilities that will address issues associated with reactor pressure vessel integrity. Ongoing research in the EPRI nuclear power group materials performance program covers a broad range of technical areas associated with RPVs. The program is structured under the following product groups; (1) management and mitigation; (2) material performance databases; (3) material condition assessment; and (4) operability assessment. Specific activities under each of theses product groups are described in this paper. (orig.)

  14. Development of PIE techniques for irradiated LWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  15. Fatigue Test of Domestic Manufactured Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    ZHONG; Wei-hua; TONG; Zhen-feng; NING; Guang-sheng; YU; Bin-tao

    2013-01-01

    The CAP1400 will be built by our country,after the self-dependent innovation work on the imported technology of AP1000,which is a 3rd generation NPP.Now,the design of CAP1400 key equipment is ongoing,and the fatigue design of the domestic manufactured key equipment,such as reactor pressure vessel(RPV),is found to be a main problem in the design work,as the fatigue data is lacked.Thus the

  16. Selection of material for building pressure vessels and chemical plants

    Energy Technology Data Exchange (ETDEWEB)

    Huppertz, P.H.; Retter, A.

    1979-06-01

    The authors give on extensive survey on the materials used in building pressure vessels and chemical plants for a temperature region of -200 to +1000/sup 0/C. The effect of various influences on the material behaviour is critically examined on the existing control plant, where the differences to foreign control are indicated. NE metals also come into consideration apart from steels, especially with low-temperature application.

  17. Thermally activated deformation of irradiated reactor pressure vessel steel

    Science.gov (United States)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  18. Composite Overwrapped Pressure Vessels (COPV) Stress Rupture Test

    Science.gov (United States)

    Russell, Richard; Flynn, Howard; Forth, Scott; Greene, Nathanael; Kezian, Michael; Varanauski, Don; Yoder, Tommy; Woodworth, Warren

    2009-01-01

    One of the major concerns for the aging Space Shuttle fleet is the stress rupture life of composite overwrapped pressure vessels (COPVs). Stress rupture life of a COPV has been defined as the minimum time during which the composite maintains structural integrity considering the combined effects of stress levels and time. To assist in the evaluation of the aging COPVs in the Orbiter fleet an analytical reliability model was developed. The actual data used to construct this model was from testing of COPVs constructed of similar, but not exactly same materials and pressure cycles as used on Orbiter vessels. Since no actual Orbiter COPV stress rupture data exists the Space Shuttle Program decided to run a stress rupture test to compare to model predictions. Due to availability of spares, the testing was unfortunately limited to one 40" vessel. The stress rupture test was performed at maximum operating pressure at an elevated temperature to accelerate aging. The test was performed in two phases. The first phase, 130 F, a moderately accelerated test designed to achieve the midpoint of the model predicted point reliability. The more aggressive second phase, performed at 160 F was designed to determine if the test article will exceed the 95% confidence interval of the model. This paper will discuss the results of this test, it's implications and possible follow-on testing.

  19. Pressure distension in leg vessels as influenced by prolonged bed rest and a pressure habituation regimen.

    Science.gov (United States)

    Eiken, Ola; Mekjavic, Igor B; Kounalakis, Stylianos N; Kölegård, Roger

    2016-06-15

    Bed rest increases pressure distension in arteries, arterioles, and veins of the leg. We hypothesized that bed-rest-induced deconditioning of leg vessels is governed by the removal of the local increments in transmural pressure induced by assuming erect posture and, therefore, can be counteracted by intermittently increasing local transmural pressure during the bed rest. Ten men underwent 5 wk of horizontal bed rest. A subatmospheric pressure (-90 mmHg) was intermittently applied to one lower leg [pressure habituation (PH) leg]. Vascular pressure distension was investigated before and after the bed rest, both in the PH and control (CN) leg by increasing local distending pressure, stepwise up to +200 mmHg. Vessel diameter and blood flow were measured in the posterior tibial artery and vessel diameter in the posterior tibial vein. In the CN leg, bed rest led to 5-fold and 2.7-fold increments (P pressure-distension and flow responses, respectively, and to a 2-fold increase in tibial vein pressure distension. In the PH leg, arterial pressure-distension and flow responses were unaffected by bed rest, whereas bed rest led to a 1.5-fold increase in venous pressure distension. It thus appears that bed-rest-induced deconditioning of leg arteries, arterioles, and veins is caused by removal of gravity-dependent local pressure loads and may be abolished or alleviated by a local pressure-habituation regimen.

  20. Composite Pressure Vessel Variability in Geometry and Filament Winding Model

    Science.gov (United States)

    Green, Steven J.; Greene, Nathanael J.

    2012-01-01

    Composite pressure vessels (CPVs) are used in a variety of applications ranging from carbon dioxide canisters for paintball guns to life support and pressurant storage on the International Space Station. With widespread use, it is important to be able to evaluate the effect of variability on structural performance. Data analysis was completed on CPVs to determine the amount of variation that occurs among the same type of CPV, and a filament winding routine was developed to facilitate study of the effect of manufacturing variation on structural response.

  1. Online Monitoring of Composite Overwrapped Pressure Vessels (COPV)

    DEFF Research Database (Denmark)

    Pereira, Gilmar Ferreira; Figueiredo, Joana; Faria, Hugo

    2015-01-01

    Composite overwrapped pressure vessels (COPV) have been increasingly pointed to as the most effective solution for high pressure storage of liquid and gaseous fluids. Reasonably high stiffness-to-weight ratios make them suitable for both static and mobile applications. However, higher operating...... pressures are sought continuously, to get higher energy densities in such storage systems, and safety aspects become critical. Thus, reliable design and test procedures are required to reduce the risks of undesired and unpredicted failures. An in-service health monitoring system may contribute to a better...... product development, design and optimization, as well as to minimize the risks and improve the public acceptance. Within the scope of developing different COPV models for a wide range of operating pressures and applications, optical fiber Bragg grating (FBG) sensors were embedded in the liner...

  2. Neutron flux reduction programs for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, C.S. [Korea Atomic Energy Research Inst. KAERI, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, B.C. [Korea Reactor Integrity Surveillance Technology KRIST, 150 Deogjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    The objective of this work is to implement various fast neutron flux reduction programs on the belt-line region of the reactor pressure vessel to reduce the increasing rate of reference temperature for pressurized thermal shock (RT PTS) for Korea Nuclear Unit 1. A pressurized thermal shock (PTS) event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in the belt-line region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Generally, the RT PTS value is continuously increasing according to the fast neutron irradiation during the reactor operation, and it can reach the screening criterion prior to the expiration of the operating license. To reduce the increasing rate of RT PTS, various neutron flux reduction programs can be implemented, which are focused on license renewal. In this paper, neutron flux reduction programs, such as low leakage loading pattern strategy, loading of neutron absorber rods, and dummy fuel assembly loading are considered for Korea Nuclear Unit 1, of which the RT PTS value of the leading material (circumferential weld) is going to reach the screening criterion in the near future. To evaluate the effects of the neutron flux reduction programs, plant and cycle specific forward neutron transport calculations for the various neutron flux reduction programs were carried out. For the analysis, all transport calculations were carried out by using the DORT 3.1 discrete ordinate code and BUGLE-96 cross-section library. (authors)

  3. STRESS ANALYSIS AND BURST PRESSURE DETERMINATION OF TWO LAYER COMPOUND PRESSURE VESSEL

    Directory of Open Access Journals (Sweden)

    HARERAM LOHAR

    2013-02-01

    Full Text Available Multilayer pressure vessel is designed to work under high-pressure condition. This paper introduces the stress analysis and the burst pressure calculation of a two-layer shrink fitted pressure vessel. In the shrink-fitting problems, considering long hollow cylinders, the plane strain hypothesis can be regarded as more natural. Generally hoops stress distribution is non-linear and sharply reduced toward the outer surface. By shrink fitting concentric shells towards the inner shells are placed in residual compression so that the initial compressive hoop stress must be relieved by internal pressure before hoop tensile stress are developed. Therefore the maximum hoop stress will be reduced, resulting more burst pressure. The analytical results of stress distribution and burst pressure is calculated and validated by ANSYS Workbench results.

  4. Delivery of cold hydrogen in glass fiber composite pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Weisberg, Andrew H.; Aceves, Salvador M.; Espinosa-Loza, Francisco; Ledesma-Orozco, Elias; Myers, Blake [Lawrence Livermore National Laboratory, Engineering, 7000 East Avenue L-792, Livermore, CA 94551 (United States)

    2009-12-15

    We are proposing to minimize hydrogen delivery cost through utilization of glass fiber tube trailers at 200 K and 70 MPa to produce a synergistic combination of container characteristics with properties of hydrogen gas: (1) hydrogen cooled to 200 K is {proportional_to}35% more compact for a small increase in theoretical storage energy (exergy); and (2) these cold temperatures (200 K) strengthen glass fibers by as much as 50%, expanding trailer capacity without the use of much more costly carbon fiber composite vessels. Analyses based on US Department of Energy H2A cost and efficiency parameters and economic methodology indicate the potential for hydrogen delivery costs below $1/kg H{sub 2}. Dispensing cold hydrogen may also allow rapid refueling without overtemperatures and overpressures which are typically as high as 25%, simplifying automotive vessel design and improving safety while potentially reducing vessel weight and cost. Based on these results, we suggest hydrogen delivery by truck with trailers carrying hydrogen gas at pressures as high as 70 MPa, cooled to approximately 200 K in glass fiber vessels. (author)

  5. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 56.13015 Section 56.13015 Mineral Resources MINE SAFETY AND HEALTH...

  6. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Repairs of boiler and pressure vessels. 78.33-1 Section... OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-1 Repairs of boiler and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer...

  7. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo tank and process pressure vessel welding. 154.650... Equipment Construction § 154.650 Cargo tank and process pressure vessel welding. (a) Cargo tank and process pressure vessel welding must meet Subpart 54.05 and Part 57 of this chapter. (b) Welding consumables...

  8. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Boilers, pressure vessels, piping and appurtenances. 167... SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-1 Boilers, pressure vessels, piping and... the following standards for boilers, pressure vessels, piping and appurtenances: (1)...

  9. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Inspection of boilers, pressure vessels, piping and...) NAUTICAL SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-5 Inspection of boilers, pressure vessels, piping and appurtenances. The inspection of boilers, pressure vessels, piping and...

  10. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Existing boilers, pressure vessels or piping systems. 50... ENGINEERING GENERAL PROVISIONS Application § 50.05-5 Existing boilers, pressure vessels or piping systems. (a) Whenever doubt exists as to the safety of an existing boiler, pressure vessel, or piping system, the...

  11. Review of roadway control in soft surrounding rock under dynamic pressure

    Institute of Scientific and Technical Information of China (English)

    侯朝炯

    2003-01-01

    The basic characteristics of the soft rock roadway under the dynamic pressure are analyzed. At the same time, the three fundamental approaches for controlling the surrounding rock are proposed, which are improving the surrounding rock strength, lowering the rock mass stress and selecting the reasonable supporting technology. The research results are elucidated, including the distribution of the surrounding rock plastic zone, the movement and damage of the surrounding rock under the dynamic pressure, controlling the floor heave through reinforcing the roadway walls and corners, the new route to develop the roadway metal supporting technique, the key theory and technique for the bolt supporting in the coal roadway, the performance and prospect of the ZKD high-water-content quick-setting material, and so on. Finally, some personally views are put forward about the roadway metal supporting, bolt supporting, new material and the stress-relief under the high stress condition.

  12. Jam proof closure assembly for lidded pressure vessels

    Science.gov (United States)

    Cioletti, Olisse C.

    1992-01-01

    An expendable closure assembly is provided for use (in multiple units) with a lockable pressure vessel cover along its rim, such as of an autoclave. This assembly is suited to variable compressive contact and locking with the vessel lid sealing gasket. The closure assembly consists of a thick walled sleeve insert for retention in the under bores fabricated in the cover periphery and the sleeve is provided with internal threading only. A snap serves as a retainer on the underside of the sleeve, locking it into an under bore retention channel. Finally, a standard elongate externally threaded bolt is sized for mating cooperation with the so positioned sleeve, whereby the location of the bolt shaft in the cover bore hole determines its compressive contact on the underlying gasket.

  13. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    Energy Technology Data Exchange (ETDEWEB)

    Hensel, S.

    2012-03-27

    Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

  14. Evaluation of Data-Logging Transducer to Passively Collect Pressure Vessel p/T History

    Science.gov (United States)

    Wnuk, Stephen P.; Le, Son; Loew, Raymond A.

    2013-01-01

    Pressure vessels owned and operated by NASA are required to be regularly certified per agency policy. Certification requires an assessment of damage mechanisms and an estimation of vessel remaining life. Since detail service histories are not typically available for most pressure vessels, a conservative estimate of vessel pressure/temperature excursions is typically used in assessing fatigue life. This paper details trial use of a data-logging transducer to passively obtain actual pressure and temperature service histories of pressure vessels. The approach was found to have some potential for cost savings and other benefits in certain cases.

  15. Structural Features and In-service Inspection of the LTNHR-200 Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    The pressure vessel of 200 MW low temperature nuclear heating reactor (LTNHR-200) is the main part of primary pressure boundary and its reasonable and reliable structural design is the key point to assure the safe operation of LTNHR-200. The double-shell pressure vessels were designed. LTNHR-200 pressure vessel meets the condition of Leak Before Break and has a relatively low failure probability. Metal containment (outer pressure vessel) has the similar features to LTNHR-200 pressure vessel. There exists no LOCA and core melting with the double vessel. The in-service inspection of the pressure vessel can be simplified greatly because of the safety and structural features of the reactor.

  16. Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    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/sup 0/C (196/sup 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 (18,700 psi).

  17. Improved fireman's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    King, H. A.; Morris, E. E.

    1973-01-01

    Prototype high pressure glass filament-wound, aluminum-lined pressurant vessels suitable for use in a fireman's compressed air breathing system were designed, fabricated, and acceptance tested in order to demonstrate the feasibility of producing such high performance, lightweight units. The 4000 psi tanks have a 60 standard cubic foot (SCF) air capacity, and have a 6.5 inch diamter, 19 inch length, 415 inch volume, weigh 13 pounds when empty, and contain 33 percent more air than the current 45 SCF (2250 psi) steel units. The current steel 60 SCF (3000 psi) tanks weigh approximately twice as much as the prototype when empty, and are 2 inches, or 10 percent shorter. The prototype units also have non-rusting aluminum interiors, which removes the hazard of corrosion, the need for internal coatings, and the possibility of rust particles clogging the breathing system.

  18. Temporal and spatial pore water pressure distribution surrounding a vertical landfill leachate recirculation well.

    Science.gov (United States)

    Kadambala, Ravi; Townsend, Timothy G; Jain, Pradeep; Singh, Karamjit

    2011-05-01

    Addition of liquids into landfilled waste can result in an increase in pore water pressure, and this in turn may increase concerns with respect to geotechnical stability of the landfilled waste mass. While the impact of vertical well leachate recirculation on landfill pore water pressures has been mathematically modeled, measurements of these systems in operating landfills have not been reported. Pressure readings from vibrating wire piezometers placed in the waste surrounding a liquids addition well at a full-scale operating landfill in Florida were recorded over a 2-year period. Prior to the addition of liquids, measured pore pressures were found to increase with landfill depth, an indication of gas pressure increase and decreasing waste permeability with depth. When liquid addition commenced, piezometers located closer to either the leachate injection well or the landfill surface responded more rapidly to leachate addition relative to those far from the well and those at deeper locations. After liquid addition stopped, measured pore pressures did not immediately drop, but slowly decreased with time. Despite the large pressures present at the bottom of the liquid addition well, much smaller pressures were measured in the surrounding waste. The spatial variation of the pressures recorded in this study suggests that waste permeability is anisotropic and decreases with depth.

  19. Retinal vessel diameter changes induced by transient high perfusion pressure

    Institute of Scientific and Technical Information of China (English)

    Yin-Ying; Zhao; Ping-Jun; Chang; Fang; Yu; Yun-E; Zhao

    2014-01-01

    ·AIM: To investigate the effects of transient high perfusion pressure on the retinal vessel diameter and retinal ganglion cells.·METHODS: The animals were divided into four groups according to different infusion pressure and infusion time(60 mm Hg-3min, 60 mm Hg-5min, 100 mm Hg-3min, 100 mm Hg-5min). Each group consisted of six rabbits. The left eye was used as the experimental eye and the right as a control. Retinal vascular diameters were evaluated before, during infusion, immediately after infusion, 5min, 10 min and 30 min after infusion based on the fundus photographs. Blood pressure was monitored during infusion. The eyes were removed after 24 h.Damage to retinal ganglion cell(RGC) was analyzed by histology.·RESULTS: Retina became whiten and papilla optic was pale during perfusion. Measurements showed significant decrease in retinal artery and vein diameter during perfusion in all of the four groups at the proximal of the edge of the optic disc. The changes were significant in the 100 mm Hg-3min group and 100 mm Hg-5min group compared with 60 mm Hg-3min group(P 1=0.025, P 2=0.000).The diameters in all the groups recovered completely after 30 min of reperfusion. The number of RGC)showed no significant changes at the IOP in 100 mm Hg with5 min compared with contralateral untreated eye(P >0.05).·CONCLUSION: Transient fluctuations during infusion lead to temporal changes of retinal vessels, which could affect the retinal blood circulation. The RGCs were not affected by this transient fluctuation. Further studies are necessary to evaluate the effect of pressure during realtime phacoemusification on retinal blood circulation.

  20. Plastic Limit Load Analysis of Cylindrical Pressure Vessels with Different Nozzle Inclination

    Science.gov (United States)

    Prakash, Anupam; Raval, Harit Kishorchandra; Gandhi, Anish; Pawar, Dipak Bapu

    2016-04-01

    Sudden change in geometry of pressure vessel due to nozzle cutout, leads to local stress concentration and deformation, decreasing its strength. Elastic plastic analysis of cylindrical pressure vessels with different inclination angles of nozzle is important to estimate plastic limit load. In the present study, cylindrical pressure vessels with combined inclination of nozzles (i.e. in longitudinal and radial plane) are considered for elastic plastic limit load analysis. Three dimensional static nonlinear finite element analyses of cylindrical pressure vessels with nozzle are performed for incremental pressure loading. The von Mises stress distribution on pressure vessel shows higher stress zones at shell-nozzle junction. Approximate plastic limit load is obtained by twice elastic slope method. Variation in limit pressure with different combined inclination angle of nozzle is analyzed and found to be distinct in nature. Reported results can be helpful in optimizing pressure vessel design.

  1. Research on reasonable winding angle of ribbons of Flat Steel Ribbon Wound Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Flat Steel Ribbon Wound Pressure Vessels (FSRWPVs) are used in many important industry areas. There is no such kind of pressure vessel exploding on operation for its reasonable structure design. Many explosion experiments on Flat Steel Ribbon Wound Pressure Vessel showed that their limited load pressure is related to the winding angle of the steel ribbons.FSRWPVs with reasonable winding angle have better security and lower cost. Reasonable angels given at the end of this paper facilitate engineering design.

  2. NASA Requirements for Ground-Based Pressure Vessels and Pressurized Systems (PVS). Revision C

    Science.gov (United States)

    Greulich, Owen Rudolf

    2017-01-01

    The purpose of this document is to ensure the structural integrity of PVS through implementation of a minimum set of requirements for ground-based PVS in accordance with this document, NASA Policy Directive (NPD) 8710.5, NASA Safety Policy for Pressure Vessels and Pressurized Systems, NASA Procedural Requirements (NPR) 8715.3, NASA General Safety Program Requirements, applicable Federal Regulations, and national consensus codes and standards (NCS).

  3. Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen and Natural Gas Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Espinosa-Loza, F; Schaffer, R; Clapper, W

    2002-05-22

    We are working on developing an alternative technology for storage of hydrogen or natural gas on light-duty vehicles. This technology has been titled insulated pressure vessels. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept either liquid fuel or ambient-temperature compressed fuel. Insulated pressure vessels offer the advantages of cryogenic liquid fuel tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for fuel liquefaction and reduced evaporative losses). The work described in this paper is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen or LNG. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining insulated pressure vessel certification.

  4. D-Zero Central Calorimeter Pressure Vessel and Vacuum Vessel Safety Notes

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, R.; Luther, R.; /Fermilab

    1990-10-25

    The relief valve and relief piping capacity was calculated to be 908 sefm air. This exceeds all relieving conditions. The vessel also has a rupture disc with a 2640 scfm air stamped capacity. In order to significantly decrease the amount of time required to fill the cryostats, it is desired to raise the setpoint of the 'operating' relief valve on the argon storage dewar to 20 psig from its existing 16 psig setting. This additional pressure increases the flow to the cryostats and will overwhelm the relief capacity if the temperature of the modules within these vessels is warm enough. Using some conservative assumptions and simple calculations within this note, the maximum average temperature that the modules within each cryostat can be at prior to filling from the storage dewar with liquid argon is at least 290 K. The average temperature of the module mass for any of the three cryostats can be as high as 290 K prior to filling that particular cryostat. This should not be confused with the average temperature of a single type or location which is useful in protecting the modules-not necessarily the vessel itself. A few modules of each type and at different elevations should be used in an average which would account for the different weights of each module. Note that at 290 K, the actual flow of argon through the relief valve and the rupture disk was under the maximum theoretical flows for each relief device. This means that the bulk temperature could actually have been raised to flow argon through the reliefs at their maximum capacity. Therefore, the temperature of 290 K is a conservative value for the calculated flow rate of 12.3 gpm. Safeguards in addition to and used in conjunction with operating procedures shall be implemented in such a way so that the above temperature limitation is not exceeded and such that it is exclusive of the programmable logic controller (PLC). One suggestion is using a toggle switch for each cryostat mounted in the PLC I/O box

  5. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  6. ACS Algorithm in Discrete Ordinates for Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Walters William

    2016-01-01

    Full Text Available The Adaptive Collision Source (ACS method can solve the Linear Boltzmann Equation (LBE more efficiently by adaptation of the angular quadrature order. This is similar to, and essentially an extension of, the first collision source method. Previously, the ACS methodology has been implemented into the TITAN discrete ordinates code, and has shown speedups of 2–4 on a simple test problem, with very little loss of accuracy (within a provided adaptive tolerance. This work examines the use of the ACS method for a more realistic problem: pressure vessel dosimetry with the VENUS-2 MOX-fuelled reactor dosimetry benchmark. The ACS method proved to be able to obtain accurate results while being approximately twice as efficient as using a constant quadrature in a standard source iteration scheme.

  7. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    Science.gov (United States)

    Ellis, Tayera

    2015-01-01

    Critical flight hardware typically undergoes a series of nondestructive evaluation methods to screen for defects before it is integrated into the flight system. Conventionally, pressure vessels have been inspected for flaws using a technique known as fluorescent dye penetrant, which is biased to inspector interpretation. An alternate method known as eddy current is automated and can detect small cracks better than dye penetrant. A new multipurpose pressure vessel scanner has been developed to perform internal and external eddy current scanning, laser profilometry, and thickness mapping on pressure vessels. Before this system can be implemented throughout industry, a probability of detection (POD) study needs to be performed to validate the system’s eddy current crack/flaw capabilities. The POD sample set will consist of 6 flight-like metal pressure vessel liners with defects of known size. Preparation for the POD includes sample set fabrication, system operation, procedure development, and eddy current settings optimization. For this, collaborating with subject matter experts was required. This technical paper details the preparation activities leading up to the POD study currently scheduled for winter 2015/2016. Once validated, this system will be a proven innovation for increasing the safety and reliability of necessary flight hardware.Additionally, testing of frangible joint requires Photon Doppler Velocimetry (PDV) and Digital Image Correlation instrumentation. There is often noise associated with PDV data, which necessitates a frequency modulation (FM) signal-to-noise pre-test. Generally, FM radio works by varying the carrier frequency and mixing it with a fixed frequency source, creating a beat frequency which is represented by audio frequency that can be heard between about 20 to 20,000 Hz. Similarly, PDV reflects a shifted frequency (a phenomenon known as the Doppler Effect) from a moving source and mixes it with a fixed source frequency, which results in

  8. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  9. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Materials Research Center; Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Cannon, N.S. [Westinghouse Hanford Co., Richland, WA (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1991-12-31

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. {Delta}USE, the difference between the USE`s of notched-only and precracked specimens, is an estimate of the crack initiation energy. {Delta}USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the {Delta}USE were found to be invariant with specimen size.

  10. Subsize specimen testing of nuclear reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.S. (Missouri Univ., Rolla, MO (United States). Materials Research Center); Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Cannon, N.S. (Westinghouse Hanford Co., Richland, WA (United States)); Hamilton, M.L. (Pacific Northwest Lab., Richland, WA (United States))

    1991-01-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, A533B. The methodology appears to be more satisfactory than the methodologies proposed earlier. USE of a notched-only specimen is partitioned into macro-crack initiation and crack propagation energies. USE of a notched and precracked specimen provides the crack propagation energy. [Delta]USE, the difference between the USE's of notched-only and precracked specimens, is an estimate of the crack initiation energy. [Delta]USE was normalized by a factor involving the dimensions of the Charpy specimen and the stress concentration factor at the notch root. The normalized values of the [Delta]USE were found to be invariant with specimen size.

  11. PRESSURE AND PRESSURE GRADIENT IN AN AXISYMMETRIC RIGID VESSEL WITH STENOSIS

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Based on an improvement of the Karman-Pohlhausen's method, using nonlinear polynomial fitting and numerical integral, the axial distributions of pressure and its gradient in an axisymmetric rigid vessel with stenosis were obtained, and the distributions related to Reynolds number and the geometry of stenotic vessel were discussed. It shows that with the increasing of stenotic degree or Reynolds number, the fluctuation of pressure and its gradient in stenotic area is intense rapidly, and negative pressure occurs subsequently in the diverging part of stenotic area. Especially when the axial range of stenosis extends, the flow of blood in the diverging part will be more obviously changed.In higher Reynolds number or heavy stenosis, theoretical calculation is mainly in accordance with past experiments.

  12. Performance Evaluation Tests of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Espinoza-Loza, F

    2002-03-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen or ambient-temperature compressed hydrogen. This flexibility results in multiple advantages with respect to compressed hydrogen tanks or low-pressure liquid hydrogen tanks. Our work is directed at verifying that commercially available aluminum-lined, fiber-wrapped pressure vessels can be safely used to store liquid hydrogen. A series of tests have been conducted, and the results indicate that no significant vessel damage has resulted from cryogenic operation. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for certification of insulated pressure vessels.

  13. Certification Testing and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Espinosa-Loza, F

    2002-05-22

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen or ambient-temperature compressed hydrogen. This flexibility results in multiple advantages with respect to compressed hydrogen tanks or low-pressure liquid hydrogen tanks. Our work is directed at verifying that commercially available aluminum-lined, fiber-wrapped pressure vessels can be safely used to store liquid hydrogen. A series of tests have been conducted, and the results indicate that no significant vessel damage has resulted from cryogenic operation. Future activities include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for certification of insulated pressure vessels.

  14. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 57.13015 Section 57.13015 Mineral Resources MINE SAFETY AND HEALTH...

  15. Performance and Certification Testing of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Garcia-Villazana, O; Espinosa-Loza, F

    2001-06-03

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH2) or ambient-temperature compressed hydrogen (CH2). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for hydrogen liquefaction and reduced evaporative losses). The work described here is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Required future tests are described that will prove that no technical barriers exist to the safe use of aluminum-fiber vessels at cryogenic temperatures. Future activities also include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining certification for insulated pressure vessels.

  16. Insulated Pressure Vessels for Vehicular Hydrogen Storage: Analysis and Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S M; Martinez-Frias, J; Garcia-Villazana, O; Espinosa-Loza, F

    2001-06-26

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (fuel flexibility, lower energy requirement for hydrogen liquefaction and reduced evaporative losses). The work described here is directed at verifying that commercially available pressure vessels can be safely used to store liquid hydrogen. The use of commercially available pressure vessels significantly reduces the cost and complexity of the insulated pressure vessel development effort. This paper describes a series of tests that have been done with aluminum-lined, fiber-wrapped vessels to evaluate the damage caused by low temperature operation. All analysis and experiments to date indicate that no significant damage has resulted. Required future tests are described that will prove that no technical barriers exist to the safe use of aluminum-fiber vessels at cryogenic temperatures. Future activities also include a demonstration project in which the insulated pressure vessels will be installed and tested on two vehicles. A draft standard will also be generated for obtaining certification for insulated pressure vessels.

  17. Upper bound analytical solution for surrounding rock pressure of shallow unsymmetrical loading tunnels

    Institute of Scientific and Technical Information of China (English)

    雷明锋; 彭立敏; 施成华; 谢友均; 谭立新

    2015-01-01

    By combining the results of laboratory model tests with relevant flow rules, the failure mode of shallow unsymmetrical loading tunnels and the corresponding velocity field were established. According to the principle of virtual power, the upper bound solution for surrounding rock pressure of shallow unsymmetrical loading tunnel was derived and verified by an example. The results indicate that the calculated results of the derived upper bound method for surrounding rock pressure of shallow unsymmetrical loading tunnels are relatively close to those of the existing “code method” and test results, which means that the proposed method is feasible. The current code method underestimates the unsymmetrical loading feature of surrounding rock pressure of shallow unsymmetrical loading tunnels, so it is unsafe; when the burial depth is less or greater than two times of the tunnel span and the unsymmetrical loading angle is less than 45°, the upper bound method or the average value of the results calculated by the upper bound method and code method respectively, is comparatively reasonable. When the burial depth is greater than two times of the tunnel span and the unsymmetrical loading angle is greater than 45°, the code method is more suitable.

  18. SAFT inspections for developing empirical database of fabrication flaws in nuclear reactor pressure vessels

    Science.gov (United States)

    Doctor, Steven R.; Schuster, George J.; Pardini, Allan F.

    1998-03-01

    The Pacific Northwest National Laboratory (PNNL) is developing a methodology for estimating the size and density distribution of fabrication flaws in U.S. nuclear reactor pressure vessels. This involves the nondestructive evaluation (NDE) of reactor pressure vessel materials and the destructive validation of the flaws found. NDE has been performed on reactor pressure vessel material made by Babcock & Wilcox and Combustion Engineering. A metallographic analysis is being performed to validate the flaw density and size distributions estimated from the 2500 indications of fabrication flaws that were detected and characterized in the very sensitive SAFT-UT (synthetic aperture focusing technique for ultrasonic testing) inspection data from the Pressure Vessel Research User Facility (PVRUF) vessel at Oak Ridge National Laboratory. Research plans are also described for expanding the work to include other reactor pressure vessel materials.

  19. Neural Network Prediction of Failure of Damaged Composite Pressure Vessels from Strain Field Data Acquired by a Computer Vision Method

    Science.gov (United States)

    Russell, Samuel S.; Lansing, Matthew D.

    1997-01-01

    This effort used a new and novel method of acquiring strains called Sub-pixel Digital Video Image Correlation (SDVIC) on impact damaged Kevlar/epoxy filament wound pressure vessels during a proof test. To predict the burst pressure, the hoop strain field distribution around the impact location from three vessels was used to train a neural network. The network was then tested on additional pressure vessels. Several variations on the network were tried. The best results were obtained using a single hidden layer. SDVIC is a fill-field non-contact computer vision technique which provides in-plane deformation and strain data over a load differential. This method was used to determine hoop and axial displacements, hoop and axial linear strains, the in-plane shear strains and rotations in the regions surrounding impact sites in filament wound pressure vessels (FWPV) during proof loading by internal pressurization. The relationship between these deformation measurement values and the remaining life of the pressure vessels, however, requires a complex theoretical model or numerical simulation. Both of these techniques are time consuming and complicated. Previous results using neural network methods had been successful in predicting the burst pressure for graphite/epoxy pressure vessels based upon acoustic emission (AE) measurements in similar tests. The neural network associates the character of the AE amplitude distribution, which depends upon the extent of impact damage, with the burst pressure. Similarly, higher amounts of impact damage are theorized to cause a higher amount of strain concentration in the damage effected zone at a given pressure and result in lower burst pressures. This relationship suggests that a neural network might be able to find an empirical relationship between the SDVIC strain field data and the burst pressure, analogous to the AE method, with greater speed and simplicity than theoretical or finite element modeling. The process of testing SDVIC

  20. Creep of A508/533 Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Richard Wright

    2014-08-01

    ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are

  1. A methodology for the control of the residual lifetimes of carbon fibre reinforced composite pressure vessels

    OpenAIRE

    Bunsell, Anthony R.; Blassiau, Sébastien; Thionnet, Alain

    2005-01-01

    International audience; Pressure vessels must be periodically proof tested. Traditional techniques for metal vessels are inapplicable for composite vessels as the latter do not break by crack propagation so that the reasoning behind the traditional testing procedures is not appropriate. Damage accumulation leading to the degradation of a composite vessel is by fibre failure. Fibres show a wide distribution in strengths and loading a composite inevitably breaks some. The method which has been ...

  2. A simplified approach for assessing the leak-before-break for the flawed pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kannan, P. [Ramagundam Super Thermal Power Station, NTPC Ltd, Jyothinagar 505215 (India); Amirthagadeswaran, K.S. [Faculty of Mechanical Engineering, Government College of Technology, Coimbatore 641013 (India); Christopher, T. [Faculty of Mechanical Engineering, Government College of Engineering, Tirunelveli 627007 (India); Nageswara Rao, B., E-mail: bnrao52@rediffmail.com [Faculty of Mechanical Engineering, School of Mechanical and Civil Sciences, K L University, Green Fields, Vaddeswaram, Guntur 522502 (India)

    2016-06-15

    Surface cracks or embedded cracks in pressure vessels under service may grow and form stable through-thickness cracks causing leak prior to failure. If this leak-before-break phenomenon takes place, then there is a possibility of preventing the vessel failure. This paper presents a simplified approach for assessing the leak-before-break or failure of the flawed pressure vessels. This approach is validated through comparison of existing test data.

  3. 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.

  4. Coordinated sensing and autonomous repair of pressure vessels and structures

    Science.gov (United States)

    Huston, Dryver R.; Hurley, David A.; Gollins, Kenneth; Gervais, Anthony

    2010-04-01

    Self-repairing structural systems can potentially improve performance ranges and lifetimes compared to those of conventional systems without self-healing capability. Self-healing materials have been used in automotive and aeronautical applications for over a century. The bulk of these systems operate by using the damage to directly initiate the repair response without any supervisory coordination. Integrating sensing and supervisory control technologies with self-healing may improve the safety and reliability of critical components and structures. This project used laboratory scale test beds to illustrate the benefit of an integrated sensing, control and self-healing system. A thermal healing polymer embedded with resistive heating wires acted as the sensing-healing material. Sensing duties were performed using an impedance, capacitance, and resistance testing device and a PC acted as the controller. As damage occurs to the polymer it is detected, located, and characterized. Based on the sensor signal, a decision is made as to whether to execute a repair and then to subsequently monitor the repair process to ensure completeness. The second demonstration was a self-sealing pressure vessel with integrated sensing and healing capability. These proof-of-concept prototypes can likely be expanded and improved with alternative sensor options, sensing-healing materials, and system architecture.

  5. 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.

  6. 77 FR 59408 - Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying...

    Science.gov (United States)

    2012-09-27

    ... SECURITY Coast Guard Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels... Coast Guard announces the availability of CG-ENG Policy Letter 04-12, ``Alternative Pressure Relief Valve Settings on Vessels Carrying Liquefied Gases in Bulk in Independent Type B and Type C...

  7. OPTIMUM AUTOFRETTAGE PRESSURE AND SHRINK-FIT COMBINATION FOR MINIMUM STRESS IN MULTILAYER PRESSURE VESSEL

    Directory of Open Access Journals (Sweden)

    S.K. Acharyaa

    2011-05-01

    Full Text Available In present work, more effective ways of decreasing the net working stress in multilayer vessel is brought into focus. Analysis of combined effect of autofrettage and shrink-fit in multi-layeredvessel is carried out. Possible sequences of assembly of autofrettage and shrink-fit in multilayered vessel have been discussed and effective sequence has been sorted out. With the increase in number of layers, sequential order of assembly increases. Optimization of thickness of each vessel for 3-layered vessel, percentage of autofrettage and amount of radial shrink-fit is carried out for all the possible sequences with the help of Genetic Algorithm. While performing optimization, thickness of each layers, autofrettage percentage and radial interference for shrink-fit is considered as design variables, whereas hoop stress throughout the thickness isobjective function. Apart from this, calculation of fatigue life for each case is studied. It is observed that all the possibilities of assembly gives approximately same behaviour under same working pressure with some exceptions.

  8. Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

    Directory of Open Access Journals (Sweden)

    Hereil Pierre-Louis

    2015-01-01

    Full Text Available Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

  9. Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

    Science.gov (United States)

    Hereil, Pierre-Louis; Plassard, Fabien; Mespoulet, Jérôme

    2015-09-01

    Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics) overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

  10. Determination of the critical buckling pressure of blood vessels using the energy approach.

    Science.gov (United States)

    Han, Hai-Chao

    2011-03-01

    The stability of blood vessels under lumen blood pressure is essential to the maintenance of normal vascular function. Differential buckling equations have been established recently for linear and nonlinear elastic artery models. However, the strain energy in bent buckling and the corresponding energy method have not been investigated for blood vessels under lumen pressure. The purpose of this study was to establish the energy equation for blood vessel buckling under internal pressure. A buckling equation was established to determine the critical pressure based on the potential energy. The critical pressures of blood vessels with small tapering along their axis were estimated using the energy approach. It was demonstrated that the energy approach yields both the same differential equation and critical pressure for cylindrical blood vessel buckling as obtained previously using the adjacent equilibrium approach. Tapering reduced the critical pressure of blood vessels compared to the cylindrical ones. This energy approach provides a useful tool for studying blood vessel buckling and will be useful in dealing with various imperfections of the vessel wall.

  11. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  12. Some Observations on Damage Tolerance Analyses in Pressure Vessels

    Science.gov (United States)

    Raju, Ivatury S.; Dawicke, David S.; Hampton, Roy W.

    2017-01-01

    AIAA standards S080 and S081 are applicable for certification of metallic pressure vessels (PV) and composite overwrap pressure vessels (COPV), respectively. These standards require damage tolerance analyses with a minimum reliable detectible flaw/crack and demonstration of safe life four times the service life with these cracks at the worst-case location in the PVs and oriented perpendicular to the maximum principal tensile stress. The standards require consideration of semi-elliptical surface cracks in the range of aspect ratios (crack depth a to half of the surface length c, i.e., (a/c) of 0.2 to 1). NASA-STD-5009 provides the minimum reliably detectible standard crack sizes (90/95 probability of detection (POD) for several non-destructive evaluation (NDE) methods (eddy current (ET), penetrant (PT), radiography (RT) and ultrasonic (UT)) for the two limits of the aspect ratio range required by the AIAA standards. This paper tries to answer the questions: can the safe life analysis consider only the life for the crack sizes at the two required limits, or endpoints, of the (a/c) range for the NDE method used or does the analysis need to consider values within that range? What would be an appropriate method to interpolate 90/95 POD crack sizes at intermediate (a/c) values? Several procedures to develop combinations of a and c within the specified range are explored. A simple linear relationship between a and c is chosen to compare the effects of seven different approaches to determine combinations of aj and cj that are between the (a/c) endpoints. Two of the seven are selected for evaluation: Approach I, the simple linear relationship, and a more conservative option, Approach III. For each of these two Approaches, the lives are computed for initial semi-elliptic crack configurations in a plate subjected to remote tensile fatigue loading with an R-ratio of 0.1, for an assumed material evaluated using NASGRO (registered 4) version 8.1. These calculations demonstrate

  13. ADDITIONAL STRESS AND FRACTURE MECHANICS ANALYSES OF PRESSURIZED WATER REACTOR PRESSURE VESSEL NOZZLES

    Energy Technology Data Exchange (ETDEWEB)

    Walter, Matthew [Structural Integrity Associates, Inc.; Yin, Shengjun [ORNL; Stevens, Gary [U.S. Nuclear Regulatory Commission; Sommerville, Daniel [Structural Integrity Associates, Inc.; Palm, Nathan [Westinghouse Electric Company, Cranberry Township, PA; Heinecke, Carol [Westinghouse Electric Company, Cranberry Township, PA

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  14. Embrittlement recovery due to annealing of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, E.D.; Wright, J.E.; Nelson, E.E. [Modeling and Computing Services, Boulder, CO (United States); Odette, G.R.; Mader, E.V. [Univ. of California, Santa Barbara, CA (United States)

    1996-03-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes.

  15. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States)

    2015-03-01

    Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

  16. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

  17. Mandibular advancement decreases pressures in the tissues surrounding the upper airway in rabbits.

    Science.gov (United States)

    Kairaitis, Kristina; Stavrinou, Rosie; Parikh, Radha; Wheatley, John R; Amis, Terence C

    2006-01-01

    The pharyngeal airway can be considered as an airway luminal shape formed by surrounding tissues, contained within a bony enclosure formed by the mandible, skull base, and cervical vertebrae. Mandibular advancement (MA), a therapy for obstructive sleep apnea, is thought to increase the size of this bony enclosure and to decrease the pressure in the upper airway extraluminal tissue space (ETP). We examined the effect of MA on upper airway airflow resistance (Rua) and ETP in a rabbit model. We studied 11 male, supine, anesthetized, spontaneously breathing New Zealand White rabbits in which ETP was measured via pressure transducer-tipped catheters inserted into the tissues surrounding the lateral (ETPlat) and anterior (ETPant) pharyngeal wall. Airflow, measured via surgically inserted pneumotachograph in series with the trachea, and tracheal pressure were recorded while graded MA at 75 degrees and 100 degrees to the horizontal was performed using an external traction device. Data were analyzed using a linear mixed-effects statistical model. We found that MA at 100 degrees increased mouth opening from 4.7 +/- 0.4 to 6.6 +/- 0.4 (SE) mm (n = 7; P < 0.004), whereas mouth opening did not change from baseline (4.0 +/- 0.2 mm) with MA at 75 degrees . MA at both 75 degrees and 100 degrees decreased mean ETPlat and ETPant by approximately 0.1 cmH2O/mm MA (n = 7-11; all P < 0.0005). However, the fall in Rua (measured at 20 ml/s) with MA was greater for MA at 75 degrees (approximately 0.03 mmH2O.ml(-1).s.mm(-1)) than at 100 degrees (approximately 0.01 mmH2O.ml(-1).s.mm(-1); P < 0.02). From these findings, we conclude that MA decreases ETP and is more effective in reducing Rua without mouth opening.

  18. Photoacoustic sample vessel and method of elevated pressure operation

    Science.gov (United States)

    Autrey, Tom; Yonker, Clement R.

    2004-05-04

    An improved photoacoustic vessel and method of photoacoustic analysis. The photoacoustic sample vessel comprises an acoustic detector, an acoustic couplant, and an acoustic coupler having a chamber for holding the acoustic couplant and a sample. The acoustic couplant is selected from the group consisting of liquid, solid, and combinations thereof. Passing electromagnetic energy through the sample generates an acoustic signal within the sample, whereby the acoustic signal propagates through the sample to and through the acoustic couplant to the acoustic detector.

  19. Review of pressurized thermal shock studies of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available In Hungary, four nuclear power units were constructed more than 30 years ago; they are operating to this day. In every unit, VVER-440 V213-type light-water cooled, light-water moderated, ressurized water reactors are in operation. Since the mid-1980s, numerous researches in the field of Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPVs have been conducted in Hungary; in all of them, the concept of structural integrity was the basis of research and development. During this time, four large PTS studies with industrial relevance have been completed in Hungary. Each used different objectives and guides, and the analysis methodology was also changing. This paper gives a comparative review of the methodologies used in these large PTS Structural Integrity Analysis projects, presenting the latest results as well

  20. Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Williams, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, B. Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Klasky, Hilda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

  1. A quick guide to API 510 certified pressure vessel inspector syllabus example questions and worked answers

    CERN Document Server

    Matthews, Clifford

    2010-01-01

    The API Individual Certification Programs (ICPs) are well established worldwide in the oil, gas, and petroleum industries. This Quick Guide is unique in providing simple, accessible and well-structured guidance for anyone studying the API 510 Certified Pressure Vessel Inspector syllabus by summarizing and helping them through the syllabus and providing multiple example questions and worked answers.Technical standards are referenced from the API 'body of knowledge' for the examination, i.e. API 510 Pressure vessel inspection, alteration, rerating; API 572 Pressure vessel inspection; API

  2. Designing of a Fleet-Leader Program for Carbon Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Murthy, Pappu L.N.; Phoenix, S. Leigh

    2009-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases on board spacecraft when mass saving is a prime requirement. Substantial weight savings can be achieved compared to all metallic pressure vessels. For example, on the space shuttle, replacement of all metallic pressure vessels with Kevlar COPVs resulted in a weight savings of about 30 percent. Mass critical space applications such as the Ares and Orion vehicles are currently being planned to use as many COPVs as possible in place of all-metallic pressure vessels to minimize the overall mass of the vehicle. Due to the fact that overwraps are subjected to sustained loads during long periods of a mission, stress rupture failure is a major concern. It is, therefore, important to ascertain the reliability of these vessels by analysis, since it is practically impossible to show by experimental testing the reliability of flight quality vessels. Also, it is a common practice to set aside flight quality vessels as "fleet leaders" in a test program where these vessels are subjected to slightly accelerated operating conditions so that they lead the actual flight vessels both in time and load. The intention of fleet leaders is to provide advanced warning if there is a serious design flaw in the vessels so that a major disaster in the flight vessels can be averted with advance warning. On the other hand, the accelerating conditions must be not so severe as to be prone to false alarms. The primary focus of the present paper is to provide an analytical basis for designing a viable fleet leader program for carbon COPVs. The analysis is based on a stress rupture behavior model incorporating Weibull statistics and power-law sensitivity of life to fiber stress level.

  3. Evaluation of fast neutron fluence for Kori Unit 2 pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Young Kyou; Lim, Mi Joung; Kim, Byoung Chul; Kim, Kyung Sik [Korea Reactor Integrity Surveillance Technology, Daejeon (Korea, Republic of)

    2011-10-15

    Unit 2 at Kori reactor has been put into operation in 1983. During 24 cycle operation, five surveillance capsules at inner vessel and three ex-vessel dosimeter at cavity both are taken out for evaluation to neutron fluence. The evaluations following the surveillance program of Kori 2 unit which are required detect and prevent degradation of safety-related structures and components of the vessel. The fast (E > 1.0 MeV) neutron fluencies are necessary to estimate the fracture toughness of the pressure vessel materials. The determination of the pressure vessel neutron fluence is based on both calculations and measurements. The fluence prediction is made with a calculation, and the measurements are used to qualify the calculational methodology. Measurement-to-calculation comparisons are used to identify biases in the calculations and to provide reliable estimates of the fluence uncertainties As shown in Fig. 1, the Kori unit 2 reactor vessel surveillance programs includes the analysis of flux dosimeters contained in capsules located on the inner vessel wall at the Beltline region (0., 15., 30. and 40. Azimuth) and Ex vessel dosimeter capsules located on the cavity at connected bid chain. In this paper, the methodologies used to perform neutron transport calculations and dosimetry evaluations are described, the results of the plant specific transport calculations are given for the beltline region of Kori Unit 2 pressure vessel and the comparisons of calculations and measurements are discussed

  4. 46 CFR 38.05-3 - Design and construction of pressure vessel type cargo tanks-TB/ALL.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Design and construction of pressure vessel type cargo... LIQUEFIED FLAMMABLE GASES Design and Installation § 38.05-3 Design and construction of pressure vessel type cargo tanks—TB/ALL. (a) Cargo tanks of pressure vessel configuration (e.g. cylindrical, spherical,...

  5. DEVELOPMENT OF ASME SECTION X CODE RULES FOR HIGH PRESSURE COMPOSITE HYDROGEN PRESSURE VESSELS WITH NON-LOAD SHARING LINERS

    Energy Technology Data Exchange (ETDEWEB)

    Rawls, G.; Newhouse, N.; Rana, M.; Shelley, B.; Gorman, M.

    2010-04-13

    The Boiler and Pressure Vessel Project Team on Hydrogen Tanks was formed in 2004 to develop Code rules to address the various needs that had been identified for the design and construction of up to 15000 psi hydrogen storage vessel. One of these needs was the development of Code rules for high pressure composite vessels with non-load sharing liners for stationary applications. In 2009, ASME approved new Appendix 8, for Section X Code which contains the rules for these vessels. These vessels are designated as Class III vessels with design pressure ranging from 20.7 MPa (3,000 ps)i to 103.4 MPa (15,000 psi) and maximum allowable outside liner diameter of 2.54 m (100 inches). The maximum design life of these vessels is limited to 20 years. Design, fabrication, and examination requirements have been specified, included Acoustic Emission testing at time of manufacture. The Code rules include the design qualification testing of prototype vessels. Qualification includes proof, expansion, burst, cyclic fatigue, creep, flaw, permeability, torque, penetration, and environmental testing.

  6. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management...

  7. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... vessels. 56.13001 Section 56.13001 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-SURFACE METAL AND NONMETAL... standards and specifications of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code....

  8. Summary of Activities for Health Monitoring of Composite Overwrapped Pressure Vessels Updated January 2014

    Science.gov (United States)

    Skow, Miles G.

    2014-01-01

    This three year project (FY12-14) will design and demonstrate the ability of new Magnetic Stress Gages for the measurement of stresses on the inner diameter of a Composite Overwrapped Pressure Vessel overwrap.

  9. Workbook for predicting pressure wave and fragment effects of exploding propellant tanks and gas storage vessels

    Science.gov (United States)

    Baker, W. E.; Kulesz, J. J.; Ricker, R. E.; Bessey, R. L.; Westine, P. S.; Parr, V. B.; Oldham, G. A.

    1975-01-01

    Technology needed to predict damage and hazards from explosions of propellant tanks and bursts of pressure vessels, both near and far from these explosions is introduced. Data are summarized in graphs, tables, and nomographs.

  10. Analysis and Design of Cryogenic Pressure Vessels for Automotive Hydrogen Storage

    Science.gov (United States)

    Espinosa-Loza, Francisco Javier

    Cryogenic pressure vessels maximize hydrogen storage density by combining the high pressure (350-700 bar) typical of today's composite pressure vessels with the cryogenic temperature (as low as 25 K) typical of low pressure liquid hydrogen vessels. Cryogenic pressure vessels comprise a high-pressure inner vessel made of carbon fiber-coated metal (similar to those used for storage of compressed gas), a vacuum space filled with numerous sheets of highly reflective metalized plastic (for high performance thermal insulation), and a metallic outer jacket. High density of hydrogen storage is key to practical hydrogen-fueled transportation by enabling (1) long-range (500+ km) transportation with high capacity vessels that fit within available spaces in the vehicle, and (2) reduced cost per kilogram of hydrogen stored through reduced need for expensive structural material (carbon fiber composite) necessary to make the vessel. Low temperature of storage also leads to reduced expansion energy (by an order of magnitude or more vs. ambient temperature compressed gas storage), potentially providing important safety advantages. All this is accomplished while simultaneously avoiding fuel venting typical of cryogenic vessels for all practical use scenarios. This dissertation describes the work necessary for developing and demonstrating successive generations of cryogenic pressure vessels demonstrated at Lawrence Livermore National Laboratory. The work included (1) conceptual design, (2) detailed system design (3) structural analysis of cryogenic pressure vessels, (4) thermal analysis of heat transfer through cryogenic supports and vacuum multilayer insulation, and (5) experimental demonstration. Aside from succeeding in demonstrating a hydrogen storage approach that has established all the world records for hydrogen storage on vehicles (longest driving range, maximum hydrogen storage density, and maximum containment of cryogenic hydrogen without venting), the work also

  11. The instability mechanics of surrounding rock-coal mass system in longwall face and the prevention of pressure bumps

    Institute of Scientific and Technical Information of China (English)

    李新元; 李英明

    2003-01-01

    According to the movement and change rules of mechanical structure of surrounding rock-coal mass system during coal excavation, the mechanism of sudden instability and damage was found out. The criterions that distinguishing the occurring of the pressure bump were put forward. This criteria have been applied successfully in the comprehensive prevent of pressure bumps in Tangshan colliery.

  12. French PWR 900 MWe pressure vessel surveillance neutron field characteristics TRIPOLI-3 calculations and experimental determination

    Energy Technology Data Exchange (ETDEWEB)

    Nimal, J.C.; Bourdet, L.; Zheng, S.H.; Vergnaud, T.; Kodeli, I. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie; Lloret, R.; Bevilacqua, A. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. des Reacteurs Experimentaux; Lefebvre, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1994-12-31

    This paper presents an overview of the studies performed by CEA and EDF in the scope of the pressure vessel surveillance of the French nuclear power plants. The power plants are equipped with surveillance capsules, attached to the thermal shield. They contain the dosimeters and vessel material specimens for monitoring the effects of irradiation on the pressure vessel material. The Monte Carlo code TRIPOLI-3 is used with two nuclear data libraries to calculate the neutron flux, the steel damage and the dosimeter reaction rates, and takes into account the results of sensitivity/uncertainty calculations. 2 figs., 7 tabs., 10 refs.

  13. Pressure vessel with impact and fire resistant coating and method of making same

    Science.gov (United States)

    DeLay, Thomas K. (Inventor)

    2005-01-01

    An impact and fire resistant coating laminate is provided which serves as an outer protective coating for a pressure vessel such as a composite overwrapped vessel with a metal lining. The laminate comprises a plurality of fibers (e.g., jute twine or other, stronger fibers) which are wound around the pressure vessel and an epoxy matrix resin for the fibers. The epoxy matrix resin including a plurality of microspheres containing a temperature responsive phase change material which changes phase in response to exposure thereof to a predetermined temperature increase so as to afford increased insulation and heat absorption.

  14. Development of a Numerical Model of Hypervelocity Impact into a Pressurized Composite Overwrapped Pressure Vessel

    Science.gov (United States)

    Garcia, M. A.; Davis, B. A.; Miller, J. E.

    2017-01-01

    As the outlook for space exploration becomes more ambitious and spacecraft travel deeper into space than ever before, it is increasingly important that propulsion systems perform reliably within the space environment. The increased reliability compels designers to increase design margin at the expense of system mass, which contrasts with the need to limit vehicle mass to maximize payload. Such are the factors that motivate the integration of high specific strength composite materials in the construction of pressure vessels commonly referred to as composite overwrapped pressure vessels (COPV). The COPV consists of a metallic liner for the inner shell of the COPV that is stiff, negates fluid permeation and serves as the anchor for composite laminates or filaments, but the liner itself cannot contain the stresses from the pressurant it contains. The compo-site-fiber reinforced polymer (CFRP) is wound around the liner using a combination of hoop (circumferential) and helical orientations. Careful consideration of wrap orientation allows the composite to evenly bear structural loading and creates the COPV's characteristic high strength to weight ratio. As the CFRP overwrap carries most of the stresses induced by pressurization, damage to the overwrap can affect mission duration, mission success and potentially cause loss-of-vehicle/loss-of-crew. For this reason, it is critical to establish a fundamental understanding of the mechanisms involved in the failure of a stressed composite such as that of the COPV. One of the greatest external threats to the integrity of a spacecraft's COPV is an impact from the meteoroid and orbital debris environments (MMOD). These impacts, even from submillimeter particles, generate extremely high stress states in the CFRP that can damage numerous fibers. As a result of this possibility, initial assumptions in survivability analysis for some human-rated NASA space-craft have assumed that any alteration of the vessel due to impact is

  15. Analytical and computational methodology to assess the over pressures generated by a potential catastrophic failure of a cryogenic pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Zamora, I.; Fradera, J.; Jaskiewicz, F.; Lopez, D.; Hermosa, B.; Aleman, A.; Izquierdo, J.; Buskop, J.

    2014-07-01

    Idom has participated in the risk evaluation of Safety Important Class (SIC) structures due to over pressures generated by a catastrophic failure of a cryogenic pressure vessel at ITER plant site. The evaluation implements both analytical and computational methodologies achieving consistent and robust results. (Author)

  16. A Review of Large-Scale Fracture Experiments Relevant to Pressure Vessel Integrity Under Pressurized Thermal Shock Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pugh, C.E.

    2001-01-29

    Numerous large-scale fracture experiments have been performed over the past thirty years to advance fracture mechanics methodologies applicable to thick-wall pressure vessels. This report first identifies major factors important to nuclear reactor pressure vessel (RPV) integrity under pressurized thermal shock (PTS) conditions. It then covers 20 key experiments that have contributed to identifying fracture behavior of RPVs and to validating applicable assessment methodologies. The experiments are categorized according to four types of specimens: (1) cylindrical specimens, (2) pressurized vessels, (3) large plate specimens, and (4) thick beam specimens. These experiments were performed in laboratories in six different countries. This report serves as a summary of those experiments, and provides a guide to references for detailed information.

  17. Stress analysis and evaluation of a rectangular pressure vessel. [For equipment for sampling Hanford tank radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Rezvani, M.A.; Ziada, H.H. (Westinghouse Hanford Co., Richland, WA (United States)); Shurrab, M.S. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1992-10-01

    This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel.

  18. Feasibility of Rectangular Concrete Pressure Vessels for Human Occupancy

    Science.gov (United States)

    1990-07-01

    initial prestressing must be sufficient to produce a resultant compressive preloadin the concrete after time dependent effects of creep in the concrete... silage . Bridge and skyscraper applications of prestressed concrete followed thereafter. Today prestressed concrete is widely used in civil engineering...for large pressurized concrete cylinders for coal gasification plants. Large high pressure, high temperature retorts were needed to convert coal into

  19. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

    2007-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  20. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

  1. Final report of the 1st ex-vessel neutron dosimetry installation and evaluations for Yonggwang unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-09-15

    This report describes a neutron fluence assessment performed for the Yonggwang unit 2 pressure vessel beltline region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During cycle 15 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Yonggwang unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 15.

  2. Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

  3. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Kim, Kwan Hyun; Hong, Joon Wha

    2007-02-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. After Cycle 22 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 23.

  4. Final report for the 3rd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Kori Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai (and others)

    2008-03-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. After Cycle 23 of reactor operation, 3rd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 24.

  5. Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  6. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

  7. Integrity analysis of reactor pressure vessels subjected to pressurized thermal shocks by XFEM

    Energy Technology Data Exchange (ETDEWEB)

    González-Albuixech, V.F., E-mail: vicente.gonzalez@psi.ch; Qian, G.; Niffenegger, M.

    2014-08-15

    Highlights: • We did fracture mechanics computations for an RPV with XFEM thermal shocks. • We introduce guidelines for using XFEM in RPV studies. • We did a comparison between FEM and XFEM results for an RPV analysis. • Some limitations of the eXtended Finite Element Methods are commented. - Abstract: The integrity of an reactor pressure vessel (RPV) related to Pressurized Thermal Shocks (PTSs) has been widely studied. However, due to the difficulties associated with the crack modeling with the 3-D finite element method (FEM), it is preferred to use models with simple geometries and crack configurations. In the last years new improved FEMs were developed which include the singularities and discontinuities and simplify the computational fracture mechanics studies. One of those methods, the eXtended Finite Element Method (XFEM) relies on the introduction of the crack effect with an enrichment of the finite element approximation space. This paper introduces the use of XFEM to the structural analysis of an RPV subjected to PTSs. The analysis compares the stress intensity factor (SIF) calculated with XFEM with results obtained by conventional FEM calculations.

  8. Detecting leaks in gas-filled pressure vessels using acoustic resonances

    Science.gov (United States)

    Gillis, K. A.; Moldover, M. R.; Mehl, J. B.

    2016-05-01

    We demonstrate that a leak from a large, unthermostatted pressure vessel into ambient air can be detected an order of magnitude more effectively by measuring the time dependence of the ratio p/f2 than by measuring the ratio p/T. Here f is the resonance frequency of an acoustic mode of the gas inside the pressure vessel, p is the pressure of the gas, and T is the kelvin temperature measured at one point in the gas. In general, the resonance frequencies are determined by a mode-dependent, weighted average of the square of the speed-of-sound throughout the volume of the gas. However, the weighting usually has a weak dependence on likely temperature gradients in the gas inside a large pressure vessel. Using the ratio p/f2, we measured a gas leak (dM/dt)/M ≈ - 1.3 × 10-5 h-1 = - 0.11 yr-1 from a 300-liter pressure vessel filled with argon at 450 kPa that was exposed to sunshine-driven temperature and pressure fluctuations as large as (dT/dt)/T ≈ (dp/dt)/p ≈ 5 × 10-2 h-1 using a 24-hour data record. This leak could not be detected in a 72-hour record of p/T. (Here M is the mass of the gas in the vessel and t is the time.)

  9. Uncertainties in risk assessment of hydrogen discharges from pressurized storage vessels at low temperatures

    DEFF Research Database (Denmark)

    Markert, Frank; Melideo, D.; Baraldi, D.

    2013-01-01

    20K) e.g. the cryogenic compressed gas storage covers pressures up to 35 MPa and temperatures between 33K and 338 K. Accurate calculations of high pressure releases require real gas EOS. This paper compares a number of EOS to predict hydrogen properties typical in different storage types. The vessel...

  10. The inclusion of weld residual stress in fracture margin assessments of embrittled nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, T.L.; Bass, B.R.; McAfee, W.J.

    1998-01-01

    Analyses were performed to determine the impact of weld residual stresses in a reactor pressure vessel (RPV) on (1) the generation of pressure temperature (P-T) curves required for maintaining specified fracture prevention margins during nuclear plant startup and shutdown, and (2) the conditional probability of vessel failure due to pressurized thermal shock (PTS) loading. The through wall residual stress distribution in an axially oriented weld was derived using measurements taken from a shell segment of a canceled RPV and finite element thermal stress analyses. The P-T curve derived from the best estimate load analysis and a t / 8 deep flaw, based on K{sub Ic}, was less limiting than the one derived from the current methodology prescribed in the ASME Boiler and Pressure Vessel Code. The inclusion of the weld residual stresses increased the conditional probability of cleavage fracture due to PTS loading by a factor ranging from 2 to 4.

  11. Could Nano-Structured Materials Enable the Improved Pressure Vessels for Deep Atmospheric Probes?

    Science.gov (United States)

    Srivastava, D.; Fuentes, A.; Bienstock, B.; Arnold, J. O.

    2005-01-01

    A viewgraph presentation on the use of Nano-Structured Materials to enable pressure vessel structures for deep atmospheric probes is shown. The topics include: 1) High Temperature/Pressure in Key X-Environments; 2) The Case for Use of Nano-Structured Materials Pressure Vessel Design; 3) Carbon based Nanomaterials; 4) Nanotube production & purification; 5) Nanomechanics of Carbon Nanotubes; 6) CNT-composites: Example (Polymer); 7) Effect of Loading sequence on Composite with 8% by volume; 8) Models for Particulate Reinforced Composites; 9) Fullerene/Ti Composite for High Strength-Insulating Layer; 10) Fullerene/Epoxy Composite for High Strength-Insulating Layer; 11) Models for Continuous Fiber Reinforced Composites; 12) Tensile Strength for Discontinuous Fiber Composite; 13) Ti + SWNT Composites: Thermal/Mechanical; 14) Ti + SWNT Composites: Tensile Strength; and 15) Nano-structured Shell for Pressure Vessels.

  12. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  13. Evaluation on Safety and Reliability of High-Pressure Vessel in Missile System

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Through theoretical analysis of reliability and simulation analysis of dispersivi of da/dN based on Monte Carlo method, the distribution function of n and c was set up.Meanwhile, the distribution of critical opening displacement ( COD ) δ, was defined by the use of coherent coefficient method, and the probabilistic model of defects assessment of military special vessel was built. Thereby the theoretical and practical fundamental research on evaluation of reliability of military high-pressure vessels was carried out.

  14. Feasibility for development of a nuclear reactor pressure vessel flaw distribution: Sensitivity analyses and NDE (nondestructive evaluation) capability

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (USA)); Kennedy, E.L.; Foulds, J.R. (Failure Analysis Associates, Inc., Menlo Park, CA (USA))

    1990-01-01

    Pressurized water reactor pressure vessels operate under US Nuclear Regulatory Commission (NRC) rules and regulatory guides that are intended to maintain a low probability of vessel failure. The NRC has also addressed neutron embrittlement of pressurized water reactor pressure vessels by imposing regulations on plant operation. Plants failing to meet the operating criteria specified by these rules and regulations are required, among other things, to analytically demonstrate fitness for service in order to continue safe operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. A fracture mechanics sensitivity study was performed to quantify the effect of the assumed flaw distribution on the predicted vessel performance under a specified pressurized thermal shock transient and to determine the critical crack size. Results of the analysis indicate that vessel performance in terms of the estimated probability of failure is very sensitive to the assumed flaw distribution. 20 refs., 3 figs., 2 tabs.

  15. Research Progress of Irradiation Embrittlement Behavior and Prediction Technology of Reactor Pressure Vessel Steel

    Institute of Scientific and Technical Information of China (English)

    YANG; Wen; TONG; Zhen-feng; NING; Guang-sheng; ZHANG; Chang-yi; BAI; Bing

    2015-01-01

    The reactor pressure vessel(RPV)is the core of the most important equipment in pressurized water reactor,and is the key equipment that cannot be replaced in nuclear power plant.The service life of RPV determines the use of nuclear power plant,and directly affects the safety and economy of nuclear power plant.Because of high temperature,high pressure and high-energy

  16. Managing Pressure Vessel Equipment as a Capital Asset.

    Science.gov (United States)

    Robinson, Glenn; Trombley, Robert; Shultes, Kenneth

    1999-01-01

    Argues the importance of treating facility pressure equipment as capital assets and discusses three steps in their management process. The following steps are discussed: understanding the condition of all major equipment; altering maintenance practices and procedures; and developing a long-term equipment strategy such as increased monitoring,…

  17. Development and application of an LWR reactor pressure vessel-specific flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. (Sandia National Labs., Albuquerque, NM (United States)); Kennedy, E.L.; Foulds, J.R. (Failure Analysis Associates, Inc., Menlo Park, CA (United States))

    1991-01-01

    Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses.

  18. Development and application of an LWR reactor pressure vessel-specific flaw distribution

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Kennedy, E.L.; Foulds, J.R. [Failure Analysis Associates, Inc., Menlo Park, CA (United States)

    1991-12-31

    Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses.

  19. Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, R.W.; Brinkman, C.R.

    1981-01-01

    Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X.

  20. 3D TRANSIENT COUPLED THERMO-ELASTIC-PLASTIC CONTACT SEALING ANALYSIS OF REACTOR PRESSURE VESSEL

    Institute of Scientific and Technical Information of China (English)

    Du Xuesong; Li Runfang; Lin Tengjiao

    2005-01-01

    Sealing analysis of sealing system in reactor pressure vessels is relevant with multiple nonlinear coupled-field effects, so even large-scale commercial finite element software cannot finish the complicated analysis. A fmite element method of 3D transient coupled thermo-elastic-plastic contact sealing analysis for reactor pressure vessels is presented, in which the surface nonlinearity,material nonlinearity, transient heat transfer nonlinearity and multiple coupled effect are taken into account and the sealing equation is coupling solved in iterative procedure. At the same time, a computational analysis program is developed, which is applied in the sealing analysis of experimental reactor pressure vessel, and the numerical results are in good coincidence with the experimental results. This program is also successful in analyzing the practical problem in engineering.

  1. Summary of Activities for Health Monitoring of Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Russell, Rick; Skow, Miles

    2013-01-01

    This three-year project (FY12-14) will design and demonstrate the ability of new Magnetic Stress Gages for the measurement of stresses on the inner diameter of a Composite Overwrapped Pressure Vessel overwrap. The sensors are being tested at White Sands Testing Facility (WSTF) where the results will be correlated with a known nondestructive technique acoustic emission. The gages will be produced utilizing Meandering Winding Magnetometer (MWM) and/or MWM array eddy current technology. The ultimate goal is to utilize this technology for the health monitoring of Composite Overwrapped Pressure Vessels for all future flight programs. The first full-scale pressurization test was performed at WSTF in June 2012. The goals of this test were to determine adaptations of the magnetic stress gauge instrumentation that would be necessary to allow multiple sensors to monitor the vessel's condition simultaneously and to determine how the sensor response changes with sensor selection and orientation. The second full scale pressurization test was performed at WSTF in August 2012. The goals of this test were to monitor the vessel's condition with multiple sensors simultaneously, to determine the viability of the multiplexing units (MUX) for the application, and to determine if the sensor responses in different orientations are repeatable. For both sets of tests the vessel was pressured up to 6,000 psi to simulate maximum operating pressure. Acoustic events were observed during the first pressurization cycle. This suggested that the extended storage period prior to use of this bottle led to a relaxation of the residual stresses imparted during auto-frettage. The pressurization tests successfully demonstrated the use of multiplexers with multiple MWM arrays to monitor a vessel. It was discovered that depending upon the sensor orientation, the frequencies, and the sense element, the MWM arrays can provide a variety of complementary information about the composite overwrapped pressure

  2. Performance features of 22-cell, 19Ah single pressure vessel nickel hydrogen battery

    Science.gov (United States)

    Rao, Gopalakrishna M.; Vaidyanathan, Hari

    1996-01-01

    Two 22-cells 19Ah Nickel-Hydrogen (Ni-H2) Single Pressure Vessel (SPV) Qual batteries, one each from EPI/Joplin and EPI/Butler, were designed and procured. The two batteries differ in the cell encapsulation technology, stack preload, and activation procedure. Both the Butler and Joplin batteries met the specified requirements when subjected to qualification testing and completed 2100 and 1300 LEO cycles respectively, with nominal performance. This paper discusses advantages, design features, testing procedures, and results of the two single pressure vessel Ni-H2 batteries.

  3. Collaborative investigations of in-service irradiated material from the Japan Power Demonstration Reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Broadhead, B.L. [Oak Ridge National Lab., TN (United States); Suzuki, M.; Kohsaka, A. [Japan Atomic Energy Research Institute, Tokai (Japan)

    1997-02-01

    There is a need to validate the results of irradiation effects research by the examination of material taken directly from the wall of a pressure vessel that has been irradiated during normal service. Just such an evaluation is currently being conducted on material from the wall of the pressure vessel from the Japan Power Demonstration Reactor (JPDR). The research is being jointly performed at the Tokai Research Establishment of the Japan Atomic Energy Research Institute (JAERI) and by the Nuclear Regulatory Commission (NRC)-funded Heavy-Section Steel Irradiation Program at the Oak Ridge National Laboratory (ORNL).

  4. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  5. Numerical simulation of premixed Hydrogen/air combustion pressure in a spherical vessel

    Directory of Open Access Journals (Sweden)

    Guo Han-yu

    2016-01-01

    Full Text Available In order to study the development process of hydrogen combustion in a closed vessel, an on-line chemical equilibrium calculator and a numerical simulation method would be used to analysis the combustion pressure and flame front of mixed gas, which based on 20L H2/air explosion experiments in spherical vessel (Crowl and Jo,2009. The results showed that, the turbulent model could reflect the process of combustion, and the error of combustion pressure by simulation is smaller than the Chemical Equilibrium Calculation. The heat loss and incomplete combustion are the main reason to cause the error.

  6. New Developments in Nickel-Hydrogen Dependent Pressure Vessel (DPV) Cell and Battery Design

    Science.gov (United States)

    Caldwell, Dwight B.; Fox, Chris L.; Miller, Lee E.

    1997-01-01

    THe Dependent Pressure Vessel (DPV) Nickel-Hydrogen (NiH2) design is being developed as an advanced battery for military and commercial, aerospace and terrestrial applications. The DPV cell design offers high specific energy and energy density as well as reduced cost, while retaining the established Individual Pressure Vessel (IPV) technology flight heritage and database. This advanced DPV design also offers a more efficient mechanical, electrical and thermal cell and battery configuration and a reduced part count. The DPV battery design promotes compact, minimum volume packaging and weight efficiency, and delivers cost and weight savings with minimal design risk.

  7. Performance features of 22-cell, 19Ah single pressure vessel nickel hydrogen battery

    Science.gov (United States)

    Rao, Gopalakrishna M.; Vaidyanathan, Hari

    1996-01-01

    Two 22-cells 19Ah Nickel-Hydrogen (Ni-H2) Single Pressure Vessel (SPV) Qual batteries, one each from EPI/Joplin and EPI/Butler, were designed and procured. The two batteries differ in the cell encapsulation technology, stack preload, and activation procedure. Both the Butler and Joplin batteries met the specified requirements when subjected to qualification testing and completed 2100 and 1300 LEO cycles respectively, with nominal performance. This paper discusses advantages, design features, testing procedures, and results of the two single pressure vessel Ni-H2 batteries.

  8. 46 CFR 50.30-10 - Class I, I-L and II-L pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class I, I-L and II-L pressure vessels. 50.30-10 Section... PROVISIONS Fabrication Inspection § 50.30-10 Class I, I-L and II-L pressure vessels. (a) Classes I, I-L and II-L pressure vessels shall be subject to shop inspection at the plant where they are...

  9. The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn; Lu, Feng; Wang, Rongshan; Huang, Ping; Liu, Xiangbin; Zhang, Guodong; Xu, Chaoliang

    2015-07-15

    Highlights: • The conservative and non-conservative assumptions in the codes were shown. • The influence of different loads on the SM was given. • The unloading effect of the cladding was studied. • A concentrated reflection of the safety was shown based on 3-D FE analyses. - Abstract: The deterministic structural integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. While the nil-ductility-transition temperature (RT{sub NDT}) parameter is widely used, the influence of fluence and temperature distributions along the thickness of the base metal wall cannot be reflected in the comparative analysis. This paper introduces the method using a structure safety margin (SM) parameter which is based on a comparison between the material toughness (the fracture initiation toughness K{sub IC} or fracture arrest toughness K{sub Ia}) and the stress intensity factor (SIF) along the crack front for the integrity analysis of a RPV subjected to PTS transients. A 3-D finite element model is used to perform fracture mechanics analyses considering both crack initiation assessment and arrest assessment. The results show that the critical part along the crack front is always the clad-base metal interface point (IP) rather than the deepest point (DP) for either crack initiation assessment or crack arrest assessment under the thermal load. It is shown that the requirement in Regulatory Guide 1.154 that ‘axial flaws with depths less than 20% of the wall thickness and all circumferential flaws should be modeled in infinite length’ may be non-conservative. As the assessment result is often poor universal for a given material, crack and transient, caution is recommended in the safety assessment, especially for the IP. The SIF reduces under the thermal or pressure load if the map cracking (MC) effect is considered. Therefore, the assumption in the ASME and RCCM codes that the cladding should be taken into account in

  10. Pressure vessel made by free forming using underwater explosion

    Directory of Open Access Journals (Sweden)

    H Iyama

    2016-03-01

    Full Text Available Explosive forming is one particular forming technique, in which, mostcommonly, water is used as the pressure transmission medium. In recentyears, we have done the development of the method which obtains anecessary form of the metal by the control of underwater shock wave actson the metal plate, without a metal die. On the other hand, the pressurevessel is required in various fields, but we think that the free forming usingthe underwater shock wave is advantageous in the production of pressurevessel of a simple spherical, ellipse, parabola shape. In this paper, we willintroduce an experiment and several numerical simulations that we carriedout for this technical development.

  11. The sensitivity of the burst performance of impact damaged pressure vessels to material strength properties

    Science.gov (United States)

    Lasn, K.; Vedvik, N. P.; Echtermeyer, A. T.

    2016-07-01

    This numerical study is carried out to improve the understanding of short-term residual strength of impacted composite pressure vessels. The relationship between the impact, created damage and residual strength is predicted by finite element (FE) analysis. The burst predictions depend largely on the strength properties used in the material models. However, it is typically not possible to measure all laminate properties on filament wound structures. Reasonable testing efforts are concentrated on critical properties, while obtaining other less sensitive parameters from e.g. literature. A parametric FE model is hereby employed to identify the critical strength properties, focusing on the cylindrical section of the pressure vessel. The model simulates an impactor strike on an empty vessel, which is subsequently pressurized until burst. Monte Carlo Simulations (MCS) are employed to investigate the correlations between strength related material parameters and the burst pressure. The simulations indicate the fracture toughness of the composite, hoop layer tensile strength and the yield stress of the PE liner as the most influential parameters for current vessel and impact configurations. In addition, the conservative variation in strength parameters is shown to have a rather moderate effect (COV ca. 7%) on residual burst pressures.

  12. Reactor Pressure Vessel Integrity Assessments with the Grizzly Aging Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin; Backman, Marie; Hoffman, William; Chakraborty, Pritam

    2015-08-01

    Grizzly is a simulation tool being developed at Idaho National Laboratory (INL) as part of the US Department of Energy’s Light Water Reactor Sustainability program to provide improved safety assessments of systems, components, and structures in nuclear power plants subjected to age-related degradation. Its goal is to provide an improved scientific basis for decisions surrounding license renewal, which would permit operation of commercial nuclear power plants beyond 60 years. Grizzly is based on INL’s MOOSE framework, which enables multiphysics simulations in a parallel computing environment. It will address a wide variety of aging issues in nuclear power plant systems, components, and structures, modelling both the aging processes and the ability of age-degraded components to perform safely. The reactor pressure vessel (RPV) was chosen as the initial application for Grizzly. Grizzly solves tightly coupled equations of heat conduction and solid mechanics to simulate the global response of the RPV to accident conditions, and uses submodels to represent regions with pre-existing flaws. Domain integrals are used to calculate stress intensity factors on those flaws. A physically based empirical model is used to evaluate material embrittlement, and is used to evaluate whether crack growth would occur. Grizzly can represent the RPV in 2D or 3D, allowing it to evaluate effects that require higher dimensionality models to capture. Work is underway to use lower length scale models of material evolution to inform engineering models of embrittlement. This paper demonstrates an application of Grizzly to RPV failure assessment, and summarizes on-going work.

  13. Development of Improved Composite Pressure Vessels for Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

    Newhouse, Norman L. [Hexagon Lincoln, Lincoln, NE (United States)

    2016-04-29

    Hexagon Lincoln started this DOE project as part of the Hydrogen Storage Engineering Center of Excellence (HSECoE) contract on 1 February 2009. The purpose of the HSECoE was the research and development of viable material based hydrogen storage systems for on-board vehicular applications to meet DOE performance and cost targets. A baseline design was established in Phase 1. Studies were then conducted to evaluate potential improvements, such as alternate fiber, resin, and boss materials. The most promising concepts were selected such that potential improvements, compared with the baseline Hexagon Lincoln tank, resulted in a projected weight reduction of 11 percent, volume increase of 4 percent, and cost reduction of 10 percent. The baseline design was updated in Phase 2 to reflect design improvements and changes in operating conditions specified by HSECoE Partners. Evaluation of potential improvements continued during Phase 2. Subscale prototype cylinders were designed and fabricated for HSECoE Partners’ use in demonstrating their components and systems. Risk mitigation studies were conducted in Phase 3 that focused on damage tolerance of the composite reinforcement. Updated subscale prototype cylinders were designed and manufactured to better address the HSECoE Partners’ requirements for system demonstration. Subscale Type 1, Type 3, and Type 4 tanks were designed, fabricated and tested. Laboratory tests were conducted to evaluate vacuum insulated systems for cooling the tanks during fill, and maintaining low temperatures during service. Full scale designs were prepared based on results from the studies of this program. The operating conditions that developed during the program addressed adsorbent systems operating at cold temperatures. A Type 4 tank would provide the lowest cost and lightest weight, particularly at higher pressures, as long as issues with liner compatibility and damage tolerance could be resolved. A Type 1 tank might be the choice if the

  14. Application of Closed Vessel Technique for the Evaluation of Burning Rates of Propellants at Low Pressures

    Directory of Open Access Journals (Sweden)

    D. Vittal

    1977-04-01

    Full Text Available Closed vessel technique has been well established for the evaluation of burning characteristics of gun, mortar and small arms propellants at high pressures of about 750 kg/cm/sup 2/ - 3000 kg/cm/sup 2/ propellants in the pressure range up to about 200 kg/cm/sup 2/ (19.6 MPa. One of the modern trends in armaments technology is development of short range, high efficiency rockets and rocket assisted projectiles where the chamber pressure are in the range of 100 kg/cm/sup 2/ - 800 kg/cm/sup 2/ (9.8 MPa-78.5 MPa. An extension of the closed vessel technique is now presented for the measurement of rates of burning of propellants in this pressure range and a few experimental results on some conventional propellants are given.

  15. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    Energy Technology Data Exchange (ETDEWEB)

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

  16. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ...)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as required otherwise by paragraph (b) of this section. Unfired steam boilers must be fitted with an efficient... § 54.15-15. Unfired steam boilers must be constructed in accordance with this part other than when...

  17. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures...

  18. Nuclear Technology. Course 30: Mechanical Inspection. Module 30-7, Pressure Vessel Inspection.

    Science.gov (United States)

    Kupiec, Chet; Espy, John

    This seventh in a series of eight modules for a course titled Mechanical Inspection is devoted to the design and fabrication of the reactor pressure vessel. The module follows a typical format that includes the following sections: (1) introduction, (2) module prerequisites, (3) objectives, (4) notes to instructor/student, (5) subject matter, (6)…

  19. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Science.gov (United States)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  20. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... life of these components. (B) The effects of localized high temperatures on degradation of the concrete... thermal annealing or to operate the nuclear power reactor following the annealing must be identified....

  1. 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

  2. Composite Overwrapped Pressure Vessels (COPV) Stress Rupture Test: Part 2. Part 2

    Science.gov (United States)

    Russell, Richard; Flynn, Howard; Forth, Scott; Greene, Nathanael; Kezirian, Michael; Varanauski, Don; Leifeste, Mark; Yoder, Tommy; Woodworth, Warren

    2010-01-01

    One of the major concerns for the aging Space Shuttle fleet is the stress rupture life of composite overwrapped pressure vessels (COPVs). Stress rupture life of a COPY has been defined as the minimum time during which the composite maintains structural integrity considering the combined effects of stress levels and time. To assist in the evaluation of the aging COPVs in the Orbiter fleet an analytical reliability model was developed. The actual data used to construct this model was from testing of COPVs constructed of similar, but not exactly same materials and pressure cycles as used on Orbiter vessels. Since no actual Orbiter COPV stress rupture data exists the Space Shuttle Program decided to run a stress rupture test to compare to model predictions. Due to availability of spares, the testing was unfortunately limited to one 40" vessel. The stress rupture test was performed at maximum operating pressure at an elevated temperature to accelerate aging. The test was performed in two phases. The first phase, 130 F, a moderately accelerated test designed to achieve the midpoint of the model predicted point reliability. A more aggressive second phase, performed at 160 F, was designed to determine if the test article will exceed the 95% confidence interval ofthe model. In phase 3, the vessel pressure was increased to above maximum operating pressure while maintaining the phase 2 temperature. After reaching enough effectives hours to reach the 99.99% confidence level of the model phase 4 testing began when the temperature was increased to greater than 170 F. The vessel was maintained at phase 4 conditions until it failed after over 3 million effect hours. This paper will discuss the results of this test, it's implications and possible follow-on testing.

  3. High-density automotive hydrogen storage with cryogenic capable pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, Salvador M.; Espinosa-Loza, Francisco; Ledesma-Orozco, Elias; Ross, Timothy O.; Weisberg, Andrew H. [Lawrence Livermore National Laboratory, P.O. Box 808, L-792, Livermore, CA 94551 (United States); Brunner, Tobias C.; Kircher, Oliver [BMW Group, Knorrstr. 147, 80788 Munich (Germany)

    2010-02-15

    LLNL is developing cryogenic capable pressure vessels with thermal endurance 5-10 times greater than conventional liquid hydrogen (LH{sub 2}) tanks that can eliminate evaporative losses in routine usage of (L)H{sub 2} automobiles. In a joint effort BMW is working on a proof of concept for a first automotive cryo-compressed hydrogen storage system that can fulfill automotive requirements on system performance, life cycle, safety and cost. Cryogenic pressure vessels can be fueled with ambient temperature compressed gaseous hydrogen (CGH{sub 2}), LH{sub 2} or cryogenic hydrogen at elevated supercritical pressure (cryo-compressed hydrogen, CcH{sub 2}). When filled with LH{sub 2} or CcH{sub 2}, these vessels contain 2-3 times more fuel than conventional ambient temperature compressed H{sub 2} vessels. LLNL has demonstrated fueling with LH{sub 2} onboard two vehicles. The generation 2 vessel, installed onboard an H{sub 2}-powered Toyota Prius and fueled with LH{sub 2} demonstrated the longest unrefueled driving distance and the longest cryogenic H{sub 2} hold time without evaporative losses. A third generation vessel will be installed, reducing weight and volume by minimizing insulation thickness while still providing acceptable thermal endurance. Based on its long experience with cryogenic hydrogen storage, BMW has developed its cryo-compressed hydrogen storage concept, which is now undergoing a thorough system and component validation to prove compliance with automotive requirements before it can be demonstrated in a BMW test vehicle. (author)

  4. Non-destructive Evaluation of Composite Pressure Vessel by Using FBG Sensors

    Institute of Scientific and Technical Information of China (English)

    HAO Jun-cai; LENG Jin-song; WEI Zhang

    2007-01-01

    In recent years, advanced composite structures are used extensively in many industries such as aerospace, aircraft, automobile,pipeline and civil engineering. Reliability and safety are crucial requirements posed by them to the advanced composite structures because of their harsh working conditions. Therefore, as a very important measure, structural health monitoring (SHM) in-service is definitely demanded for ensuring their safe working in-situ. In this paper, fiber Bragg grating (FBG) sensors are surface-mounted on the hoop and in the axial directions of a FRP pressure vessel to monitor the strain status during its pressurization. The experimental results show that the FBG sensors could be used to monitor the strain development and determine the ultimate failure strain of the composite pressure vessel.

  5. Standard practice for examination of seamless, gas-filled, steel pressure vessels using angle beam ultrasonics

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice describes a contact angle-beam shear wave ultrasonic technique to detect and locate the circumferential position of longitudinally oriented discontinuities and to compare the amplitude of the indication from such discontinuities to that of a specified reference notch. This practice does not address examination of the vessel ends. The basic principles of contact angle-beam examination can be found in Practice E 587. Application to pipe and tubing, including the use of notches for standardization, is described in Practice E 213. 1.2 This practice is appropriate for the ultrasonic examination of cylindrical sections of gas-filled, seamless, steel pressure vessels such as those used for the storage and transportation of pressurized gasses. It is applicable to both isolated vessels and those in assemblies. 1.3 The practice is intended to be used following an Acoustic Emission (AE) examination of stacked seamless gaseous pressure vessels (with limited surface scanning area) described in Test Met...

  6. Sensitivity coefficients for the stochastic estimation of the radiation damage to the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez, C.M.; Hernandez Valle, S. [Centro de Investigaciones Tecnologicas, Nucleares y Ambientales, La Habana (Cuba). E-mail: calvarez@ctn.isctn.edu.cu; svalle@ctn.isctn.edu.cu

    2000-07-01

    The construction of the sensitivity matrix in the case of the vessel radiation damage estimation by Monte Carlo techniques poses new problems related to the uncertainties of the obtained responses. In the case of deterministic calculations, the sensitivity coefficient obtention is a straightforward procedure based on the perturbation formalism through the calculation of the adjoint fluxes. In the paper an alternative procedure implementation based on the differential operator method is described with the modifications needed to the used HEXANN-EVALU code for the response estimations in the VVER-440 pressure vessel. (author)

  7. Dosimetry analyses of the Ringhals 3 and 4 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kulesza, J.A.; Fero, A.H. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Rouden, J.; Green, E.L. [Vattenfall/Ringhals AB, 432 85 Vaeroebacka (Sweden)

    2011-07-01

    A comprehensive series of neutron dosimetry measurements consisting of surveillance capsules, reactor pressure vessel cladding samples, and ex-vessel neutron dosimetry has been analyzed and compared to the results of three-dimensional, cycle-specific neutron transport calculations for the Ringhals Unit 3 and Unit 4 reactors in Sweden. The comparisons show excellent agreement between calculations and measurements. The measurements also demonstrate that it is possible to perform retrospective dosimetry measurements using the {sup 93}Nb (n,n') {sup 93m}Nb reaction on samples of 18-8 austenitic stainless steel with only trace amounts of elemental niobium. (authors)

  8. Study on calculation of rock pressure for ultra-shallow tunnel in poor surrounding rock and its tunneling procedure

    Institute of Scientific and Technical Information of China (English)

    Xiaojun Zhou; Jinghe Wang; Bentao Lin

    2014-01-01

    A computational method of rock pressure applied to an ultra-shallow tunnel is presented by key block theory, and its mathematical formula is proposed according to a mechanical tunnel model with super-shallow depth. Theoretical analysis shows that the tunnel is subject to asymmetric rock pressure due to oblique topography. The rock pressure applied to the tunnel crown and sidewall is closely related to the surrounding rock bulk density, tunnel size, depth and angle of oblique ground slope. The rock pressure applied to the tunnel crown is much greater than that to the sidewalls, and the load applied to the left side-wall is also greater than that to the right sidewall. Mean-while, the safety of the lining for an ultra-shallow tunnel in strata with inclined surface is affected by rock pressure and tunnel support parameters. Steel pipe grouting from ground surface is used to consolidate the unfavorable surrounding rock before tunnel excavation, and the reinforcing scope is proposed according to the analysis of the asymmetric load induced by tunnel excavation in weak rock with inclined ground surface. The tunneling procedure of bench cut method with pipe roof protection is still discussed and carried out in this paper according to the special geological condition. The method and tunneling procedure have been successfully utilized to design and drive a real expressway tunnel. The practice in building the super-shallow tunnel has proved the feasibility of the calculation method and tunneling procedure presented in this paper.

  9. Fracture assessment of the Oskarshamn 1 reactor pressure vessel under cold over-pressurization

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, I. [DNV Technical Consulting AB, Stockholm (Sweden)

    2001-03-01

    The major motivation of this study was to develop a methodology for fracture assessment of surface defects in the 01 reactor pressure vessel under cold loading scenarios, particularly the cold over-pressurization event. According to a previous study, the FENIX project, the cold over-pressurization of the O1 reactor is a limiting loading case, as the ductile/brittle transition temperature (RT{sub NDT}) of certain welds in the O1 beltline region may be over 100 deg C at the-end-of-life condition. The FENIX project gave values of the acceptable and critical crack depth to be equal to the thickness of the cladding layer (about 6 mm) under this load case using the ASME K{sub Ic} reference curve methodology. This study is aimed to develop a methodology to give a more precise fracture assessment of the O1 reactor under cold loading scenarios. Some of the main objectives of this study have been as below: To prepare a material which can simulate the mechanical properties and RT{sub NDT} of the O1 reactor at the end-of-life conditions. To conduct a fracture mechanics test program to cover the essential influencing factors, such as crack geometry (shallow and deep cracks) and loading condition (uniaxial and biaxial) on the cleavage fracture toughness. To perform fracture mechanics analyses to identify a suitable methodology for assessment of the experimental results. To study the responses of engineering fracture assessment methods to the experimental results from the clad specimens. To propose a fracture assessment procedure for determination of the acceptable and critical flaw sizes in the 01 reactor under the cold loading events. A test program consisted of experiments on standard SEN(B) specimens and clad beams, containing surface cracks was conducted during the course of this project. A total of nine clad beams and clad cruciform specimens were tested under uniaxial and biaxial loading. The test material is reactor steel of type A 508 Grade B, which is specially heat

  10. Methodological developments in the field of structural integrity analyses of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available Buildings, structures and systems of large scale and high value (e.g. conventional and nuclear power plants, etc. are designed for a certain, limited service lifetime. If the standards and guidelines of the time are taken into account during the design process, the resulting structures will operate safely in most cases. However, in the course of technical history there were examples of unusual, catastrophic failures of structures, even resulting in human casualties. Although the concept of Structural Integrity first appeared in industrial applications only two-three decades ago, its pertinence has been growing higher ever since. Four nuclear power generation units have been constructed in Hungary, more than 30 years ago. In every unit, VVER-440 V213 type light-water cooled, light-water moderated, pressurized water reactors are in operation. Since the mid-1980s, Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPV have been conducted in Hungary, where the concept of structural integrity was the basis of research and development. In the first part of the paper, a short historic overview is given, where the origins of the Structural Integrity concept are presented, and the beginnings of Structural Integrity in Hungary are summarized. In the second part, a new conceptual model of Structural Integrity is introduced. In the third part, a brief description of the VVER-440 V213 type RPV and its surrounding primary system is presented. In the fourth part, a conceptual model developed for PTS Structural Integrity Analyses is explained.

  11. 46 CFR 54.01-2 - Adoption of division 1 of section VIII of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... Boiler and Pressure Vessel Code. 54.01-2 Section 54.01-2 Shipping COAST GUARD, DEPARTMENT OF HOMELAND... division 1 of section VIII of the ASME Boiler and Pressure Vessel Code. (a) Pressure vessels shall be designed, constructed, and inspected in accordance with section VIII of the ASME Boiler and Pressure...

  12. Analysis of deterministic and statistical approaches to fatigue crack growth in pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, Alexandre S.; Melo, P.F. Frutuoso e [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear. E-mail: frutuoso@lmn.con.ufrj.br

    2000-07-01

    This work presents three approaches to the fatigue crack growth process in steel pressure vessels as applied to failure probability calculation. In the Thomson's methodology, the crack growth is the term that represents the mechanical behavior which along the time will take the pressure vessel to a structural failure. The first result of failure probability will be obtained considering a deterministic approach, since the crack growth laws are of a deterministic nature. This approach will provide a reference value. Next, two statistical approaches will be performed based on the fact that fatigue crack growth is a random phenomenon. One of them takes into account only the variability of experimental data, proposing a distribution function to represent the failure process. The other, the stochastic approach, considers the random nature of crack growth along time, by performing the randomization of a crack growth law. The solution of this stochastic equation is a transition distribution function fitted to experimental data. (author)

  13. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M. [Nuclear Research Institute Rez plc (Czech Republic); Steele, L.E. [Chief Scientific Investigator of the Programme, Springfield, VA (United States)

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  14. Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs). Corrected Copy, Aug. 25, 2014

    Science.gov (United States)

    Prosser, William H.

    2014-01-01

    In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

  15. Retrospective Dosimetry of Vver 440 Reactor Pressure Vessel at the 3RD Unit of Dukovany Npp

    Science.gov (United States)

    Marek, M.; Viererbl, L.; Sus, F.; Klupak, V.; Rataj, J.; Hogel, J.

    2009-08-01

    Reactor pressure vessel (RPV) residual lifetime of the Czech VVER-440 is currently monitored under Surveillance Specimens Programs (SSP) focused on reactor pressure vessel materials. Neutron fluence in the samples and its distribution in the RPV are determined by a combination of calculation results and the experimental data coming from the reactor dosimetry measurements both in the specimen containers and in the reactor cavity. The direct experimental assessment of the neutron flux density incident onto RPV and neutron fluence for the entire period of nuclear power plant unit operation can be based on the evaluation of the samples taken from the inner RPV cladding. The Retrospective Dosimetry was also used at Dukovany NPP at its 3rd unit after the 18th cycle. The paper describes methodology, experimental setup for sample extraction, measurement of activities, and the determination of the neutron flux and fluence averaged over the samples.

  16. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  17. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  18. Evaluation of Acoustic Emission NDE of Kevlar Composite Over Wrapped Pressure Vessels

    Science.gov (United States)

    Horne, Michael R.; Madaras, Eric I.

    2008-01-01

    Pressurization and failure tests of small Kevlar/epoxy COPV bottles were conducted during 2006 and 2007 by Texas Research Institute Austin, Inc., at TRI facilities. This is a report of the analysis of the Acoustic Emission (AE) data collected during those tests. Results of some of the tests indicate a possibility that AE can be used to track the stress-rupture degradation of COPV vessels.

  19. Placement of trans-sternal wires according to an ellipsoid pressure vessel model of sternal forces.

    Science.gov (United States)

    Casha, Aaron R; Manché, Alex; Gauci, Marilyn; Camilleri-Podesta, Marie-Therese; Schembri-Wismayer, Pierre; Sant, Zdenka; Gatt, Ruben; Grima, Joseph N

    2012-03-01

    Dehiscence of median sternotomy wounds remains a clinical problem. Wall forces in thin-walled pressure vessels can be calculated by membrane stress theory. An ellipsoid pressure vessel model of sternal forces is presented together with its application for optimal wire placement in the sternum. Sternal forces were calculated by computational simulation using an ellipsoid chest wall model. Sternal forces were correlated with different sternal thicknesses and radio-density as measured by computerized tomography (CT) scans of the sternum. A comparison of alternative placement of trans-sternal wires located either at the levels of the costal cartilages or the intercostal spaces was made. The ellipsoid pressure vessel model shows that higher levels of stress are operative at increasing chest diameter (P cartilage levels when compared with adjacent intercostal spaces. This results in a decrease of average sternal stress from 438 kPa at the intercostal space level to 338 kPa at the costal cartilage level (P = 0.003). Biomechanical modelling suggests that placement of trans-sternal wires at the thicker bone and more radio-dense level of the costal cartilages will result in reduced stress.

  20. Robinson 2 reactor vessel: pressurized thermal shock analysis for a small-break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Marston, T.; Griesbach, T.; Chao, J.; Chexal, B.; Norris, D.; Nickell, B.; Layman, B.

    1984-08-01

    A best-estimate Pressurized Thermal Shock (PTS) analysis was performed for a three-inch diameter hot-leg small-break loss-of-coolant accident for the Robinson 2 plant. This plant specific analysis was performed using EPRI's linked set of codes for PTS analysis. The analysis shows that with the H.B. Robinson 2 reactor pressure vessel, a hot-leg small-break loss-of-coolant accident does not pose a significant health or safety concern to the public for at least 40 years of operation.

  1. Environmental crack-growth behavior of high strength pressure vessel alloys

    Science.gov (United States)

    Forman, R. G.

    1975-01-01

    Results of sustained-load environmental crack growth threshold tests performed on six spacecraft pressure vessel alloys are presented. The alloys were Inconel 718, 6Al-4V titanium, A-286 steel, AM-350 stainless steel, cryoformed AISI 301 stainless steel; and cryoformed AISI 304L steel. The test environments for the program were air, pressurized gases of hydrogen, oxygen, nitrogen, and carbon dioxide, and liquid environments of distilled water, sea water, nitrogen tetroxide, hydrazine, aerozine 50, monomethyl hydrazine, and hydrogen peroxide. Surface flaw type specimens were used with flaws located in both base metal and weld metal.

  2. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

  3. NUMERICAL PREDICTION OF HIGHER SELF-PRESSURIZATION RATES IN A TYPICAL STORAGE VESSEL

    Directory of Open Access Journals (Sweden)

    HARI KRISHNA RAJ

    2012-07-01

    Full Text Available Self-pressurization, as a result of vaporization can occur in many scientific and technical applications like cryogenic storage tanks, pressurized water reactors etc. Predictions of both the pressurization and vaporization rates are vital in defining design requirements conforming to the tank’s maximum working pressure andexpected liquid losses. Predicting precisely the highly transient interface phenomenon due to mass transfer coupled with phase change due to evaporation is the major challenge encountered in modeling selfpressurization. The recent improvements of the multiphase flow modeling in the ANSYS FLUENT code make it now possible to simulate these mechanisms in detail without the need of user defined functions. The volume-of-fluid (VOF method in conjunction with evaporation–condensation mass transfer model has been used here. In this paper we are extending the proven capability of VOF model for predicting higher selfpressurization rates due to phase change in storage vessels.

  4. Pressure vessels dossier restoration according to NR-13 requirements; Enquadramento de vasos de pressao a norma NR-13

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Jose L. [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil); Goncalves, Osorio C. [PETROBRAS Transporte S.A. (TRANSPETRO), Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Pressure vessels are static pressurized equipment typical in oil industry facilities. In TRANSPETRO terminals and stations as well as in the whole PETROBRAS, these equipment can be found in the form of condenser accumulators, separators, heat exchangers, storage spheres and others. Because they work sustaining pressure and, many times flammable fluids, pressure vessels have a reasonable potential for hazard. For this reason, the NR-13 regulation was created. It deals with the safety in maintenance, operation and inspection of pressure vessels and boilers. During the compliance to the NR- 13 rules, a problem usually found is the lack of documents for different reasons. In this case, the NR-13 obligates the owner to recreate the vessel documentation under the responsibility of a chartered professional. This paper presents a case study where NR-13 rules were conformed by tasks involving documentation reconstruction based on information collected by means of inspection and tests performed on the field. (author)

  5. Automatic computerized measurement of retinal blood vessels with adaptive tracking algorithm and association with blood pressure

    Directory of Open Access Journals (Sweden)

    Andrej Ikica

    2007-05-01

    Full Text Available Background: To validate an automatic computer-based method for measuring the caliber of retinal blood vessels and use it to determine the effects of arterial hypertension on the calibers of these vessels and on their ratio.Methods: 295 patients with increased blood pressure were analyzed. All arterioles and venules located in the area between one half and one disc diameter from the optic disc margin were measured with the computer based program. These measurements were combined to provide the average diameters of retinal arterioles and venules and the association with blood pressure was analyzed. The arteriole-to-venule ratio (AVR was also calculated.Results: The average arteriolar diameter of patients who had hypertension from 5 to 15 years was 89.311 μm. Patients with hypertension for more than 15 years they had value of 79.276 μm. Average venular diameters were very similar in both groups (103.319 μm vs. 101.392 μm. We noticed differences in average arteriolar diameter between control group and hypertonic patients who had hypertension for more than 15 years (92.083 μm vs. 79.276 μm. Venular differences were minimal. The average of retinal venules in control group was 106.029 μm, in patients with hypertension for more than 15 years it was 101.392 μm.Conclusions: Using a computer-assisted method to measure retinal vessel diameter we found out that the diameter of retinal arterioles narrowed with blood pressure level. Our findings demonstrate a relation between presence and severity of hypertension and retinal diameter. Diameter of retinal venules hardly changed. Such relationship was similar with men and women. Fully automated system for analyzing retinal vessels is simple to use, quick and reliable.

  6. Numerical investigations on pressurized AL-composite vessel response to hypervelocity impacts: Comparison between experimental works and a numerical code

    Directory of Open Access Journals (Sweden)

    Mespoulet Jérôme

    2015-01-01

    Full Text Available Response of pressurized composite-Al vessels to hypervelocity impact of aluminum spheres have been numerically investigated to evaluate the influence of initial pressure on the vulnerability of these vessels. Investigated tanks are carbon-fiber overwrapped prestressed Al vessels. Explored internal air pressure ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from experiments (Xray radiographies, particle velocity measurement and post-mortem vessels have been compared to numerical results given from LS-DYNA ALE-Lagrange-SPH full coupling models. Simulations exhibit an under estimation in term of debris cloud evolution and shock wave propagation in pressurized air but main modes of damage/rupture on the vessels given by simulations are coherent with post-mortem recovered vessels from experiments. First results of this numerical work are promising and further simulation investigations with additional experimental data will be done to increase the reliability of the simulation model. The final aim of this crossed work is to numerically explore a wide range of impact conditions (impact angle, projectile weight, impact velocity, initial pressure that cannot be explore experimentally. Those whole results will define a rule of thumbs for the definition of a vulnerability analytical model for a given pressurized vessel.

  7. Numerical investigations on pressurized AL-composite vessel response to hypervelocity impacts: Comparison between experimental works and a numerical code

    Science.gov (United States)

    Mespoulet, Jérôme; Plassard, Fabien; Hereil, Pierre-Louis

    2015-09-01

    Response of pressurized composite-Al vessels to hypervelocity impact of aluminum spheres have been numerically investigated to evaluate the influence of initial pressure on the vulnerability of these vessels. Investigated tanks are carbon-fiber overwrapped prestressed Al vessels. Explored internal air pressure ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from experiments (Xray radiographies, particle velocity measurement and post-mortem vessels) have been compared to numerical results given from LS-DYNA ALE-Lagrange-SPH full coupling models. Simulations exhibit an under estimation in term of debris cloud evolution and shock wave propagation in pressurized air but main modes of damage/rupture on the vessels given by simulations are coherent with post-mortem recovered vessels from experiments. First results of this numerical work are promising and further simulation investigations with additional experimental data will be done to increase the reliability of the simulation model. The final aim of this crossed work is to numerically explore a wide range of impact conditions (impact angle, projectile weight, impact velocity, initial pressure) that cannot be explore experimentally. Those whole results will define a rule of thumbs for the definition of a vulnerability analytical model for a given pressurized vessel.

  8. Pressure vessel deformation under in-vessel retention condition%熔融物堆内滞留条件下压力容器变形

    Institute of Scientific and Technical Information of China (English)

    温爽; 李铁萍; 李聪新; 高新力

    2016-01-01

    熔融物堆内滞留(In-Vessel Retention, IVR)已经成为第三代反应堆一项关键的严重事故缓解策略,而压力容器外部冷却(External Reactor Vessel Cooling, ERVC)技术则是保证IVR得以成功实施的关键。当发生堆芯熔化时,高温熔融物对压力容器(Reactor Pressure Vessel, RPV)下封头的热冲击会导致RPV壁面和由其构成的外部冷却通道的形状发生变化,使局部传热恶化,进而造成IVR的失效。因此,有必要对IVR条件下RPV壁面的变形进行研究。本文利用有限元软件ANSYS对RPV进行了几何建模、温度场分析和力学场分析。结果表明,在RPV外部实现冷却、内部实现泄压的前提下,壁面变形为13.85−18.75 mm。在1 MPa内压的作用下,高温蠕变会使壁面变形随时间增大,但其增量有限。热膨胀是造成壁面变形的主要因素。%Background: In-vessel retention (IVR) has become an important severe accident mitigation strategy for advanced light water reactor in recent years. The successful implementation of IVR depends on the external reactor vessel cooling (ERVC) technique. In case of core melt, the bottom head of reactor pressure vessel (RPV) becomes deformed due to the thermal impacts of high temperature, and causes the narrowing of external coolant channel which is the gap between pressure vessel outer wall and insulation layer. This phenomenon could lead to local heat transfer deterioration and then causes the failure of IVR.Purpose: The aim of this paper is to analyze the deformation of reactor pressure vessel under IVR condition.Methods: The thermal and mechanical calculations of reactor pressure vessel are performed by using the finite element methods. This work can be divided into two steps. The first step is the evaluation of the thermal field of RPV, and the second step is the calculation of stress and displacement of RPV based on its temperature fields.Results: The result shows that the maximum vertical

  9. Cleavage Fracture Modeling of Pressure Vessels under Transient Thermo-Mechanical Loading

    Energy Technology Data Exchange (ETDEWEB)

    Qian, Xudong [National University of Singapore; Dodds, Robert [University of Illinois; Yin, Shengjun [ORNL; Bass, Bennett Richard [ORNL

    2008-02-01

    The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) will combine nonlinear analyses of crack-front response with stochastic treatments of crack size, shape, orientation, location, material properties and thermal-pressure transients. The projected computational demands needed to support stochastic approaches with detailed 3-D, nonlinear stress analyses of vessels containing defects appear well beyond current and near-term capabilities. In the interim, 2-D models become appealing to approximate certain classes of critical flaws in RPVs, and have computational demands within reach for stochastic frameworks. The present work focuses on the capability of 2-D models to provide values for the Weibull stress fracture parameter with accuracy comparable to those from very detailed 3-D models. Weibull stress approaches provide one route to connect nonlinear vessel response with fracture toughness values measured using small laboratory specimens. The embedded axial flaw located in the RPV wall near the cladding-vessel interface emerges from current linear-elastic, stochastic investigations as a critical contributor to the conditional probability of initiation. Three different types of 2-D models reflecting this configuration are subjected to a thermal-pressure transient characteristic of a critical pressurized thermal shock event. The plane-strain, 2-D models include: the modified boundary layer (MBL) model, the middle tension (M(T)) model, and the 2-D RPV model. The 2-D MBL model provides a high quality estimate for the Weibull stress but only in crack-front regions with a positive T-stress. For crack-front locations with low constraint (T-stress < 0), the M(T) specimen provides very accurate Weibull stress values but only for pressure load acting alone on the RPV. For RPVs under a combined thermal-pressure transient, Weibull stresses computed from the 2-D RPV model demonstrate close agreement with those computed from the

  10. Cleavage Fracture Modeling of Pressure Vessels Under Transient Thermo-Mechanical Loading

    Energy Technology Data Exchange (ETDEWEB)

    Qian, Xudong [National University of Singapore; Dodds, Robert [University of Illinois; Yin, Shengjun [ORNL; Bass, Bennett Richard [ORNL

    2008-01-01

    Abstract The next generation of fracture assessment procedures for nuclear reactor pressure vessels (RPVs) will combine nonlinear analyses of crack-front response with stochastic treatments of crack size, shape, orientation, location, material properties and thermal-pressure transients. The projected computational demands needed to support stochastic approaches with detailed 3-D, nonlinear stress analyses of vessels containing defects appear well beyond current and near-term capabilities. In the interim, 2-D models be-come appealing to approximate certain classes of critical flaws in RPVs, and have computational demands within reach for stochastic frameworks. The present work focuses on the capability of 2-D models to provide values for the Weibull stress fracture parameter with accuracy comparable to those from very detailed 3-D models. Weibull stress approaches provide one route to connect nonlinear vessel response with fracture toughness values measured using small laboratory specimens. The embedded axial flaw located in the RPV wall near the cladding-vessel interface emerges from current linear-elastic, stochastic investigations as a critical contributor to the conditional probability of initiation. Three different types of 2-D models reflecting this configuration are subjected to a thermal-pressure transient characteristic of a critical pressurized thermal shock event. The plane-strain, 2-D models include: the modified boundary layer (MBL) model, the middle tension (M(T)) model, and the 2-D RPV model. The 2-D MBL model provides a high quality estimate for the Weibull stress but only in crack-front regions with a positive T-stress. For crack-front locations with low constraint (T-stress < 0), the M(T) specimen provides very accurate Weibull stress values but only for pressure load acting alone on the RPV. For RPVs under a combined thermal-pressure transient, Weibull stresses computed from the 2-D RPV model demonstrate close agreement with those computed from

  11. Fabrication Flaw Density and Distribution In Repairs to Reactor Pressure Vessel and Piping Welds

    Energy Technology Data Exchange (ETDEWEB)

    GJ Schuster, FA Simonen, SR Doctor

    2008-04-01

    The Pacific Northwest National Laboratory is developing a generalized fabrication flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in U.S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This report describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. The relevance of construction records is established for describing fabrication processes and product forms. An analysis of these records shows there was a significant change in repair frequency over the years when these components were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance. Fabrication flaws in repairs are characterized using optimized-access, high-sensitivity nondestructive ultrasonic testing. Flaw characterizations are then validated by other nondestructive evaluation techniques and complemented by destructive testing.

  12. Monitoring Composite Material Pressure Vessels with a Fiber-Optic/Microelectronic Sensor System

    Science.gov (United States)

    Klimcak, C.; Jaduszliwer, B.

    1995-01-01

    We discuss the concept of an integrated, fiber-optic/microelectronic distributed sensor system that can monitor composite material pressure vessels for Air Force space systems to provide assessments of the overall health and integrity of the vessel throughout its entire operating history from birth to end of life. The fiber optic component would include either a semiconductor light emitting diode or diode laser and a multiplexed fiber optic sensing network incorporating Bragg grating sensors capable of detecting internal temperature and strain. The microelectronic components include a power source, a pulsed laser driver, time domain data acquisition hardware, a microprocessor, a data storage device, and a communication interface. The sensing system would be incorporated within the composite during its manufacture. The microelectronic data acquisition and logging system would record the environmental conditions to which the vessel has been subjected to during its storage and transit, e.g., the history of thermal excursions, pressure loading data, the occurrence of mechanical impacts, the presence of changing internal strain due to aging, delamination, material decomposition, etc. Data would be maintained din non-volatile memory for subsequent readout through a microcomputer interface.

  13. Are Retinal Vessels Calibers Influenced by Blood Pressure Measured at the Time of Retinography Acquisition?

    Directory of Open Access Journals (Sweden)

    Sandra C Fuchs

    Full Text Available Retinal arterial narrowing is associated with higher office blood pressure (BP and ambulatory blood pressure monitoring, and increased incidence of cardiovascular disease, but it is still unknown if the vessel caliber is associated with BP measured at the time of retinography acquisition.Retinal arteriolar and venular calibers were measured by the microdensitometric method in 448 patients with hypertension. Participants underwent 24-hours ambulatory blood pressure (24-h ABP monitoring simultaneously with the retinography acquisition. Association between arteriolar and venular calibers with increase of 10 mmHg in the mean 24-hours, daily, and nightly BP, and with BP measured at the time of retinography, was evaluated by ANOVA and multivariate analyses.Mean 24-hours, daytime and nighttime systolic and diastolic BP were inversely associated with the arteriolar caliber, but not with the venular caliber. Arteriolar caliber decreased -0.8 (95% CI -1.4 to -0.2 μm per 10-mmHg increase in 24-hours mean systolic BP, adjusted for age, gender, fellow vessel, and duration of hypertension (P = 0.01. The corresponding decreasing in arteriolar caliber by 10 mmHg of increasing in mean diastolic BP was -1.1 μm (-2.0 to -0.2, P = 0.02. The decrease of arteriolar caliber by the same increasing of BP measured at the time of retinography was lower and not statistically significant, particularly for mean diastolic BP and outer arterioles calibers: -1.0 (-1.8 to -0.2 μm in the daytime BP average versus -0.3 (-0.9 to 0.3 at the moment of retinography acquisition.These findings suggest that the caliber of arteriolar retinal vessels in patients with uncontrolled hypertension are not significantly influenced by blood pressure measured at the time of retinography acquisition.

  14. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  15. Proceedings of the DOE/SNL/EPRI sponsored Reactor Pressure Vessel Thermal Annealing Workshop. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Carter, R.G. [ed.] [Electric Power Research Institute, Charlotte, NC (United States)

    1994-09-01

    The purpose of the Reactor Pressure vessel Thermal Annealing Workshop was to provide a forum for US utilities and interested parties to discuss relevant experience and issues and identify potential solutions/approaches related to: An understanding of the potential benefits of thermal annealing for US commercial reactors; on-going technical research activities; technical aspects of a generic, full-scale, in-place vessel annealing demonstration; and the impact of economic, regulatory, and technical issues on the application of thermalannealingtechnology to US plants. Experts from the international nuclear reactor community were brought together to discuss issues regarding application of thermal annealing technology in the US and identify the steps necessary to commercialize this technology for US reactors. These proceedings contain all presentation materials discussed during the Workshop. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  16. Proceedings of the DOE/SNL/EPRI sponsored Reactor Pressure Vessel Thermal Annealing Workshop. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Carter, R.G. [ed.] [Electric Power Research Institute, Charlotte, NC (United States)

    1994-09-01

    The purpose of the Reactor Pressure Vessel Thermal Annealing Workshop was to provide a forum for US utilities and interested parties to discuss relevant experience and issues and identify potential solutions/approaches related to: (1) an understanding of the potential benefits of thermal annealing for US commercial reactors; (2) on-going technical research activities; (3) technical aspects of a generic, full-scale, in-place vessel annealing demonstration; and (4) the impact of economic, regulatory, and technical issues on the application of thermal annealing technology to US plants. Experts from the international nuclear reactor community were brought together to discuss issues regarding application of thermal annealing technology in the US and identify the steps necessary to commercialize this technology for US reactors. These proceedings contain all presentation materials discussed during the Workshop. This document, Volume 2, contains sections 10 through 13, Individual papers have been cataloged separately.

  17. Application of the small punch test to reactor pressure vessel integrity

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI Nondestructive Evaluation Center, Charlotte, NC (United States); Viswanathan, R. [EPRI, Palo Alto, CA (United States); Foulds, J.R. [Failure Analysis Associates, Inc., Menlo Park, CA (United States)

    1998-07-01

    Based on prior success with fossil plant steels, EPRI is investigating the feasibility of applying the Small Punch test to determine the fracture toughness (K{sub ic}) of irradiated reactor pressure vessel (RPV) materials. The small punch test specimen is sufficiently small to alleviate future surveillance material availability concerns, as well as provide a means of direct vessel material interrogation by non-disruptive miniature sample removal and testing. A limited series of small punch tests on unirradiated and irradiated RPV steel materials has shown that the method can be used to estimate ductile-to-brittle transition temperatures and to determine the material fracture toughness (K{sub lc}, J{sub lc}). The results to date are described and the experimental difficulties that need to be resolved in achieving valid results are identified. (authors)

  18. Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J

    1983-01-01

    Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development.

  19. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.; RECINIELLO,R.N.; HU,J.P.; RORER,D.C.

    2002-08-18

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex{trademark} polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup.

  20. Are Retinal Vessels Calibers Influenced by Blood Pressure Measured at the Time of Retinography Acquisition?: e0136678

    National Research Council Canada - National Science Library

    Sandra C Fuchs; Helena M Pakter; Marcelo K Maestri; Marina Beltrami-Moreira; Miguel Gus; Leila B Moreira; Manuel M Oliveira; Flavio D Fuchs

    2015-01-01

    ...) and ambulatory blood pressure monitoring, and increased incidence of cardiovascular disease, but it is still unknown if the vessel caliber is associated with BP measured at the time of retinography acquisition...

  1. UT digitized data processing for in service inspection of pressurized water reactor vessels

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, F.; Hernandez, L. [Intercontrole, Rungis (France); Paradis, L. [CEA/CEREM, 91191, Gifs/Yvette cedex (France)

    1998-03-01

    Pressurized reactor vessels in France have been examined from the inside with ultrasonic focused transducers since the very first inspection (Gagnor and Levy (1993)). The developments carried out in collaboration with the French Atomic Energy Commission (CEA) to improve the characterization of flaws detected in the body of the vessels or in the nozzles, in the vicinity of the inner or the outer surfaces now have application throughout the CIVAMIS software. The processing modules of CIVAMIS, which are implemented on site since 1994 and used by INTERCONTROLE during the in service inspections of the French PWR vessels, allow full characterization of these specific flaws. The first module is devoted to the characterization of defects located near the outer surface of the vessel or the bottom head welds (OSD module). It includes the modeling software MEPHISTOMIS which predicts the echoes coming from the interaction between the ultrasonic beam and the defects. The second module of CIVAMIS (inner surface defect module called ISD), applied to the analysis of flaws expected near the inner surface of the vessels, has been used during performance demonstration exercises on qualification mock-ups, and also on-site in five expert appraisals since its qualification in 1995. The third module available on the system has beendeveloped and qualified for the analysis of flaws likely to appear near the inner surface of the no zzles. This module, named `undercladding crack defect` (UCD) module, provides the operators with a set of pre-defined processing configurations well adapted to the characteristics of the transducers. (orig.) 11 refs.

  2. Swedish Work on Brittle-Fracture Problems in Nuclear Reactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M.

    1966-03-15

    After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO

  3. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    Energy Technology Data Exchange (ETDEWEB)

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  4. Damage Control Plan for International Space Station Recharge Tank Assembly Composite Overwrapped Pressure Vessel

    Science.gov (United States)

    Cook, Anthony J.

    2011-01-01

    As NASA has retired the Space Shuttle Program, a new method of transporting compressed gaseous nitrogen and oxygen needed to be created for delivery of these crucial life support resources to the International Space Station (ISS). One of the methods selected by NASA includes the use of highly pressurized, unprotected Recharge Tank Assemblies (RTAs) utilizing Composite Overwrapped Pressure Vessels (COPVs). A COPV consists of a thin liner wrapped with a fiber composite and resin or epoxy. It is typically lighter weight than an all metal pressure vessel of similar volume and therefore provides a higher-efficiency means for gas storage. However COPVs are known to be susceptible to damage resulting from handling, tool drop impacts, or impacts from other objects. As a result, a comprehensive Damage Control Plan has been established to mitigate damage to the RTA COPV throughout its life cycle. The DCP is intended to evaluate and mitigate defined threats during manufacturing, shipping and handling, test, assembly level integration, shipment while pressurized, launch vehicle integration and mission operations by defining credible threats and methods for preventing potential damage while still maintaining the primary goal of resupplying ISS gas resources. A comprehensive threat assessment is performed to identify all threats posed to the COPV during the different phases of its lifecycle. The threat assessment is then used as the basis for creating a series of general inspection, surveillance and reporting requirements which apply across all phases of the COPV's life, targeted requirements only applicable to specific work phases and a series of training courses for both ground personnel and crew aboard the ISS. A particularly important area of emphasis deals with creating DCP requirements for a highly pressurized, large and unprotected RTA COPV for use during Inter Vehicular Activities (IVA) operations in the micro gravity environment while supplying pressurized gas to the

  5. The criteria of fracture in the case of the leak of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Habil; Ziliukas, A.

    1997-04-01

    In order to forecast the break of the high pressure vessels and the network of pipes in a nuclear reactor, according to the concept of leak before break of pressure vessels, it is necessary to analyze the conditions of project, production, and mounting quality as well as of exploitation. It is also necessary to evaluate the process of break by the help of the fracture criteria. In the Ignalina Nuclear Power Plant of, in Lithuania, the most important objects of investigation are: the highest pressure pipes, made of Japanese steel 19MN5 and having an anticorrosive austenitic: coal inside, the pipes of distribution, which arc made of 08X1810T steel. The steel of the network of pipes has a quality of plasticity: therefore the only criteria of fragile is impossible to apply to. The process of break would be best described by the universal criteria of elastic - plastic fracture. For this purpose the author offers the criterion of the double parameter.

  6. Standard practice for examination of seamless, Gas-Filled, pressure vessels using acoustic emission

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 This practice provides guidelines for acoustic emission (AE) examinations of seamless pressure vessels (tubes) of the type used for distribution or storage of industrial gases. 1.2 This practice requires pressurization to a level greater than normal use. Pressurization medium may be gas or liquid. 1.3 This practice does not apply to vessels in cryogenic service. 1.4 The AE measurements are used to detect and locate emission sources. Other nondestructive test (NDT) methods must be used to evaluate the significance of AE sources. Procedures for other NDT techniques are beyond the scope of this practice. See Note 1. Note 1—Shear wave, angle beam ultrasonic examination is commonly used to establish circumferential position and dimensions of flaws that produce AE. Time of Flight Diffraction (TOFD), ultrasonic examination is also commonly used for flaw sizing. 1.5 The values stated in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only. 1.6 This standa...

  7. Fundamental study of failure mechanisms of pressure vessels under thermo-mechanical cycling in multiphase environments

    Science.gov (United States)

    Penso Mula, Jorge Antonio

    Cracking and bulging in welded and internally lined pressure vessels that work in thermal-mechanical cycling services have been well known problems in the petrochemical, power and nuclear industries. Published literature and industry surveys show that similar problems have been occurring during the last 50 years. Understanding the causes of cracking and bulging would lead to improvements in the reliability of these pressure vessels. This study attempts to add information required for improving the knowledge and fundamental understanding of these problems. Cracking and bulging, most often in the weld areas, commonly experienced in delayed coking units (e.g. coke drums) in oil refineries are typical examples. The coke drum was selected for this study because of the existing field experience and past industrial investigation results that were available to serve as the baseline references for the analytical studies performed for this dissertation. Another reason for selecting the delayed coking units for this study was due to their high economical yields. Shutting down these units would cause a high negative economic impact on the refinery operations. Several failure mechanisms were hypothesized. The finite element method was used to analyze these significant variables and to verify the hypotheses. In conclusion, a fundamental explanation of the occurrence of bulging and cracking in pressure vessels in multiphase environments has been developed. Several important factors have been identified, including the high convection coefficient of the boiling layer during filling and quenching, the mismatch in physical, thermal and mechanical properties in the dissimilar weld of the clad plates and process conditions such as heating and quenching rate and warming time. Material selection for coke drums should consider not only fatigue strength but also corrosion resistance at high temperatures and low temperatures. Cracking occurs due to low cycle fatigue and corrosion. The FEA

  8. The biomechanics of erections: two- versus one-compartment pressurized vessel modeling of the penis.

    Science.gov (United States)

    Mohamed, Ahmed M; Erdman, Arthur G; Timm, Gerald W

    2010-12-01

    Previous biomechanical models of the penis simulated penile erections utilizing 2D geometry, simplified 3D geometry or made inaccurate assumptions altogether. These models designed the shaft of the penis as a one-compartment pressurized vessel fixed at one end when in reality it is a two-compartment pressurized vessel in which the compartments diverge as they enter the body and are fixed at two separate anatomic sites. This study utilizes the more anatomically correct two-compartment penile model to investigate erectile function. Simplified 2D and 3D models of the erect penis were developed using the finite element method with varying anatomical considerations for analyzing structural stresses, axial buckling, and lateral deformation. This study then validated the results by building and testing corresponding physical models. Finally, a more complex and anatomically accurate model of the penis was designed and analyzed. When subject to a lateral force of 0.5 N, the peak equivalent von Mises (EVM) stress in the two-compartment model increased by about 31.62%, while in the one-compartment model, the peak EVM stress increased by as high as 70.11%. The peak EVM stress was 149 kPa in the more complex and anatomically accurate penile model. When the perforated septum was removed, the peak EVM stress increased to 455 kPa. This study verified that there is significant difference between modeling the penis as a two- versus a one-compartment pressurized vessel. When subjected to external forces, a significant advantage was exhibited by two corporal based cavernosal bodies separated by a perforated septum as opposed to one corporal body. This is due to better structural integrity of the tunica albuginea when subjected to external forces.

  9. Preliminary investigation of an ultrasound method for estimating pressure changes in deep-positioned vessels

    Science.gov (United States)

    Olesen, Jacob Bjerring; Villagomez-Hoyos, Carlos Armando; Traberg, Marie Sand; Chee, Adrian J. Y.; Yiu, Billy Y. S.; Ho, Chung Kit; Yu, Alfred C. H.; Jensen, Jørgen Arendt

    2016-04-01

    This paper presents a method for measuring pressure changes in deep-tissue vessels using vector velocity ultrasound data. The large penetration depth is ensured by acquiring data using a low frequency phased array transducer. Vascular pressure changes are then calculated from 2-D angle-independent vector velocity fields using a model based on the Navier-Stokes equations. Experimental scans are performed on a fabricated flow phantom having a constriction of 36% at a depth of 100 mm. Scans are carried out using a phased array transducer connected to the experimental scanner, SARUS. 2-D fields of angle-independent vector velocities are acquired using directional synthetic aperture vector flow imaging. The obtained results are evaluated by comparison to a 3-D numerical simulation model with equivalent geometry as the designed phantom. The study showed pressure drops across the constricted phantom varying from -40 Pa to 15 Pa with a standard deviation of 32%, and a bias of 25% found relative to the peak simulated pressure drop. This preliminary study shows that pressure can be estimated non-invasively to a depth that enables cardiac scans, and thereby, the possibility of detecting the pressure drops across the mitral valve.

  10. Evaluation of the risk of pressure vessel failure due to errors in the manufacturing process

    Energy Technology Data Exchange (ETDEWEB)

    Marriott, D.L. (Illinois Univ., Urbana); Beyers, C.J.E.

    1983-01-01

    A pilot study based on the construction of a small pressure vessel is described together with a general safety assessment strategy. It is concluded that the application of safety assessment to a manufacturing process is feasible, and that useful information regarding the improvement of control of such processes can be so derived. The most important contribution is considered to be the development of a systematic strategy for the identification of potential material failure mechanisms in a given manufacturing process. An alternative probability measure is proposed for evaluating risk. This is a subjective measure of the uncertainty of the available information, and has implications for the possible quantification of quality assurance activities.

  11. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

    1999-12-01

    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  12. Review of the Palisades pressure vessel accumulated fluence estimate and of the least squares methodology employed

    Energy Technology Data Exchange (ETDEWEB)

    Griffin, P.J.

    1998-05-01

    This report provides a review of the Palisades submittal to the Nuclear Regulatory Commission requesting endorsement of their accumulated neutron fluence estimates based on a least squares adjustment methodology. This review highlights some minor issues in the applied methodology and provides some recommendations for future work. The overall conclusion is that the Palisades fluence estimation methodology provides a reasonable approach to a {open_quotes}best estimate{close_quotes} of the accumulated pressure vessel neutron fluence and is consistent with the state-of-the-art analysis as detailed in community consensus ASTM standards.

  13. Evaluation of dynamic fracture toughness for Yong Gwang unit 5 reactor pressure vessel materials (Baseline Tests)

    Energy Technology Data Exchange (ETDEWEB)

    Chi Se Hwan; Kim, Joo Hag; Hong, Jun Hwa; Kwon, Sun Chil; Lee, Bong Sang [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    The dynamic fracture toughness (K{sub d}) of intermediate shell and its weld in SA 508 CI. 3 Yong Gwang 5 reactor pressure vessel was determined and evaluated. Precracked thirty six Charpy specimens were tested by using an instrumented impact tester. The purpose of present work is to evaluate and confirm the un-irradiated dynamic fracture toughness and to provide pre-irradiation baseline data for future evaluation on dynamic fracture toughness change during operation. 18 refs., 5 figs., 5 tabs. (Author)

  14. Towards economic design of a pressure vessel made of duplex stainless steel

    OpenAIRE

    Veljkovic, Milan, ed. lit.; Gozzi, Jonas

    2005-01-01

    The feasibility of manufacturing pressure vessels from duplex stainless steel (grade EN 1.4462, duplex 2205: 0.02%C, 22%Cr, 5.7%Ni, 3.1%Mo, 0.17%N) was evaluated by measurements of the mechanical properties of the parent metal and welds, and by finite element modelling. The tensile stress-strain properties of 4 mm sheet, and of the weld metal and the HAZ were measured in the transverse and rolling directions. Behaviour under biaxial stress was studied by pressurising circular sheet specimens ...

  15. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  16. Inspection of dissimilar metal welds in reactor pressure vessels in Spain

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J.R.; Regidor, J.J.; Pelaez, J.A.; Serrano, P. [Tecnatom, S.A., San Sebastian de los Reyes, Madrid (Spain)

    2011-07-01

    MRP-139 recommendations for inspection of dissimilar metal (DM) welds in PWR vessels were launched in the last years in the USA. Basically, it increases the frequency of the examinations in these type of welds, with major emphasis in the hot loops, adding one intermediate inspection at the ten years interval in outlet nozzles. The spanish nuclear power plants (NPP's) have begun the implementation of this type of inspections on the vessel nozzles DM welds. As this type of inspections could have an impact in the critical path duration of the outage, it is necessary the use of a mechanical equipment able to examine the nozzles DM welds in a short vessel occupation time (VOT) with high quality, qualified techniques and minimum requirements of the refuelling platform. Tecnatom undertook the design and development of a new more advanced equipment, named TENIS-DM, for implementing the reactor pressure vessel (RPV) nozzles examination. This equipment was designed in order to accomplish the stringent requirements and the updated examination techniques; it was used for the inspection of the DM welds of Asco 1 NPP inlet and outlet nozzles in March 2011. Examination techniques and procedures were qualified through the GRUVAL validation program, based on ENIC methodology. Mechanical scanner was equipped with a large number of examination probes, and TV cameras -for visual inspection and also for monitoring the ultrasonic inspections. A remote operated submarine was also used to give support to the operational personnel during the manipulation of the equipment and its movements from one nozzle to the others. During two months before the inspection, tests of the complete inspection system were made on a nozzle mock-up installed in a 4 meters deep well at Tecnatom's facilities; this scenario was also used during the training sessions of the inspection crew. The defined technical and practical objectives were achieved: use of qualified techniques and minimal impact on the

  17. Blood Pressure Control in Aging Predicts Cerebral Atrophy Related to Small-Vessel White Matter Lesions

    Directory of Open Access Journals (Sweden)

    Kyle C. Kern

    2017-05-01

    Full Text Available Cerebral small-vessel damage manifests as white matter hyperintensities and cerebral atrophy on brain MRI and is associated with aging, cognitive decline and dementia. We sought to examine the interrelationship of these imaging biomarkers and the influence of hypertension in older individuals. We used a multivariate spatial covariance neuroimaging technique to localize the effects of white matter lesion load on regional gray matter volume and assessed the role of blood pressure control, age and education on this relationship. Using a case-control design matching for age, gender, and educational attainment we selected 64 participants with normal blood pressure, controlled hypertension or uncontrolled hypertension from the Northern Manhattan Study cohort. We applied gray matter voxel-based morphometry with the scaled subprofile model to (1 identify regional covariance patterns of gray matter volume differences associated with white matter lesion load, (2 compare this relationship across blood pressure groups, and (3 relate it to cognitive performance. In this group of participants aged 60–86 years, we identified a pattern of reduced gray matter volume associated with white matter lesion load in bilateral temporal-parietal regions with relative preservation of volume in the basal forebrain, thalami and cingulate cortex. This pattern was expressed most in the uncontrolled hypertension group and least in the normotensives, but was also more evident in older and more educated individuals. Expression of this pattern was associated with worse performance in executive function and memory. In summary, white matter lesions from small-vessel disease are associated with a regional pattern of gray matter atrophy that is mitigated by blood pressure control, exacerbated by aging, and associated with cognitive performance.

  18. The vascular Ca2+-sensing receptor regulates blood vessel tone and blood pressure.

    Science.gov (United States)

    Schepelmann, M; Yarova, P L; Lopez-Fernandez, I; Davies, T S; Brennan, S C; Edwards, P J; Aggarwal, A; Graça, J; Rietdorf, K; Matchkov, V; Fenton, R A; Chang, W; Krssak, M; Stewart, A; Broadley, K J; Ward, D T; Price, S A; Edwards, D H; Kemp, P J; Riccardi, D

    2016-02-01

    The extracellular calcium-sensing receptor CaSR is expressed in blood vessels where its role is not completely understood. In this study, we tested the hypothesis that the CaSR expressed in vascular smooth muscle cells (VSMC) is directly involved in regulation of blood pressure and blood vessel tone. Mice with targeted CaSR gene ablation from vascular smooth muscle cells (VSMC) were generated by breeding exon 7 LoxP-CaSR mice with animals in which Cre recombinase is driven by a SM22α promoter (SM22α-Cre). Wire myography performed on Cre-negative [wild-type (WT)] and Cre-positive (SM22α)CaSR(Δflox/Δflox) [knockout (KO)] mice showed an endothelium-independent reduction in aorta and mesenteric artery contractility of KO compared with WT mice in response to KCl and to phenylephrine. Increasing extracellular calcium ion (Ca(2+)) concentrations (1-5 mM) evoked contraction in WT but only relaxation in KO aortas. Accordingly, diastolic and mean arterial blood pressures of KO animals were significantly reduced compared with WT, as measured by both tail cuff and radiotelemetry. This hypotension was mostly pronounced during the animals' active phase and was not rescued by either nitric oxide-synthase inhibition with nitro-l-arginine methyl ester or by a high-salt-supplemented diet. KO animals also exhibited cardiac remodeling, bradycardia, and reduced spontaneous activity in isolated hearts and cardiomyocyte-like cells. Our findings demonstrate a role for CaSR in the cardiovascular system and suggest that physiologically relevant changes in extracellular Ca(2+) concentrations could contribute to setting blood vessel tone levels and heart rate by directly acting on the cardiovascular CaSR.

  19. Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hoffman, William [Univ. of Idaho, Moscow, ID (United States); Sen, Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geometry, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodology for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically

  20. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  1. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Apu, E-mail: asarkar@barc.gov.in; Kumawat, Bhupendra K.; Chakravartty, J.K.

    2015-07-15

    The cyclic stress–strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain–stress relationships and the strain–life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  2. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  3. Digital Cellular Solid Pressure Vessels: A Novel Approach for Human Habitation in Space

    Science.gov (United States)

    Cellucci, Daniel; Jenett, Benjamin; Cheung, Kenneth C.

    2017-01-01

    It is widely assumed that human exploration beyond Earth's orbit will require vehicles capable of providing long duration habitats that simulate an Earth-like environment - consistent artificial gravity, breathable atmosphere, and sufficient living space- while requiring the minimum possible launch mass. This paper examines how the qualities of digital cellular solids - high-performance, repairability, reconfigurability, tunable mechanical response - allow the accomplishment of long-duration habitat objectives at a fraction of the mass required for traditional structural technologies. To illustrate the impact digital cellular solids could make as a replacement to conventional habitat subsystems, we compare recent proposed deep space habitat structural systems with a digital cellular solids pressure vessel design that consists of a carbon fiber reinforced polymer (CFRP) digital cellular solid cylindrical framework that is lined with an ultra-high molecular weight polyethylene (UHMWPE) skin. We use the analytical treatment of a linear specific modulus scaling cellular solid to find the minimum mass pressure vessel for a structure and find that, for equivalent habitable volume and appropriate safety factors, the use of digital cellular solids provides clear methods for producing structures that are not only repairable and reconfigurable, but also higher performance than their conventionally manufactured counterparts.

  4. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  5. Determination of Pressure Profile During Closed-Vessel Test Through CFD Simulation

    Institute of Scientific and Technical Information of China (English)

    Ahmed Bougamra∗; Huilin Lu

    2016-01-01

    Two⁃phase flow modeling of solid propellants has great potential for simulating and predicting the ballistic parameters in closed vessel tests as well as in guns. This paper presents a numerical model describing the combustion of a solid propellant in a closed chamber and takes into account what happens in such two⁃phase, unsteady, reactive⁃flow systems. The governing equations are derived in the form of coupled, non⁃linear axisymmetric partial differential equations. The governing equations with customized parameters are implemented into Ansys Fluent 14�5. The presented solutions predict the pressure profile inside the closed chamber. The results show that the present code adequately predicts the pressure⁃time history. The numerical results are in agreement with the experimental results. Some discussions are given regarding the effect of the grain shape and the sensitivity of these predictions to the initial pressure of the solid propellant bed. The study demonstrates the capability of using the present model implemented into Fluent, to simulate the combustion of solid propellants in a closed vessel and, eventually, the interior ballistic process in guns.

  6. CNG transport by ship with FRP pressure vessels access to east coast gas

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, S. [Trans Ocean Gas Inc., St. John' s, NL (Canada)

    2005-07-01

    This paper discussed the Trans Ocean Gas (TOG) method for transporting compressed natural gas (CNG). CNG transportation offers an alternative method for transporting stranded natural gas to existing markets and for creating new natural gas markets that are not feasible for liquefied natural gas (LNG) or pipelines. Trans Ocean Gas Inc. (TOG) modified an existing fibre reinforced plastic (FRP) pressure vessel technology to safely store CNG on a ship. The newly developed containment system has proven to overcome all the deficiencies of steel-based systems. TOG patented the containment system and will license its use to owners of stranded gas and shipping service providers around the world. The CNG systems will be built and assembled throughout facilities in Atlantic Canada. FRP pressure vessels have been proven safe and reliable through critical applications in the national defense, aerospace, and natural gas vehicle industries. They are light-weight, highly reliable, have very safe failure modes, are corrosion resistant, and have excellent low temperature characteristics. Under TOG's scheme, natural gas can be stored at two thirds the density of LNG without costly processing. TOG's proposed design and testing of a CNG system was reviewed in detail. figs.

  7. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Science.gov (United States)

    Sarkar, Apu; Kumawat, Bhupendra K.; Chakravartty, J. K.

    2015-07-01

    The cyclic stress-strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain-stress relationships and the strain-life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  8. A stochastic-bayesian model for the fracture probability of PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, Alexandre S.; Duran, Jorge Alberto R., E-mail: afrancisco@metal.eeimvr.uff.br, E-mail: duran@metal.eeimvr.uff.br [Universidade Federal Fluminense (UFF), Volta Redonda, RJ (Brazil). Dept. de Engenharia Mecanica

    2013-07-01

    Fracture probability of pressure vessels containing cracks can be obtained by methodologies of easy understanding, which require a deterministic treatment, complemented by statistical methods. However, more accurate results are required, methodologies need to be better formulated. This paper presents a new methodology to address this problem. First, a more rigorous methodology is obtained by means of the relationship of probability distributions that model crack incidence and nondestructive inspection efficiency using the Bayes' theorem. The result is an updated crack incidence distribution. Further, the accuracy of the methodology is improved by using a stochastic model for the crack growth. The stochastic model incorporates the statistical variability of the crack growth process, combining the stochastic theory with experimental data. Stochastic differential equations are derived by the randomization of empirical equations. From the solution of this equation, a distribution function related to the crack growth is derived. The fracture probability using both probability distribution functions is in agreement with theory, and presents realistic value for pressure vessels. (author)

  9. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  10. 46 CFR 53.01-3 - Adoption of section IV of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section IV of the ASME Boiler and Pressure...) MARINE ENGINEERING HEATING BOILERS General Requirements § 53.01-3 Adoption of section IV of the ASME Boiler and Pressure Vessel Code. (a) Heating boilers shall be designed, constructed, inspected,...

  11. 46 CFR 52.01-2 - Adoption of section I of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section I of the ASME Boiler and Pressure...) MARINE ENGINEERING POWER BOILERS General Requirements § 52.01-2 Adoption of section I of the ASME Boiler and Pressure Vessel Code. (a) Main power boilers and auxiliary boilers shall be designed,...

  12. A continuum damage analysis of hydrogen attack in a 2.25Cr–1Mo pressure vessel

    NARCIS (Netherlands)

    Burg, M.W.D. van der; Giessen, E. van der; Tvergaard, V.

    1998-01-01

    A micromechanically based continuum damage model is presented to analyze the stress, temperature and hydrogen pressure dependent material degradation process termed hydrogen attack, inside a pressure vessel. Hydrogen attack (HA) is the damage process of grain boundary facets due to a chemical reacti

  13. Assessment of the advantages of a residual heat removal system inside the reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Gautier, G.M.

    1995-12-31

    In the framework of research on diversified means for removing the residual heat from pressurized water reactors, the CEA is studying a passive system called RRP (Refroidissement du Reacteur au Primaire, or primary circuit cooling system), which includes integrated heat-exchangers and a layout of the internal structures so as to obtain convection from the primary circuit inside the vessel, whatever the state of the loops. This system is operational for all primary circuit temperatures and pressures, as well as for a wide range of conditions: it is independent of the state of the loops, even if the volume of water in the primary circuit is small, it is compatible with either a passive or an active operation mode, and compatible with any other decay heat removal systems. An evaluation is presented here of the performance of the RRP system in the event of a small primary circuit break in a totally passive operation mode without the intervention of another system. The results of this evaluation show the interest of such a system: a clear increase of the time-delay for the implementation of a low pressure safety injection system, no need for the use of a high pressure safety injection system. (author). 4 refs., 7 figs., 1 tab.

  14. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Matjaž Leskovar

    2016-02-01

    Full Text Available A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

  15. Effect of Heat Flux on Creep Stresses of Thick-Walled Cylindrical Pressure Vessels

    Directory of Open Access Journals (Sweden)

    Mosayeb Davoudi Kashkoli

    2014-06-01

    Full Text Available Assuming that the thermo-creep response of the material is governed by Norton’s law, an analytical solution is presented for the calculation of time-dependent creep stresses and displacements of homogeneous thick-walled cylindrical pressure vessels. For the stress analysis in a homogeneous pressure vessel, having material creep behavior, the solutions of the stresses at a time equal to zero (i.e. the initial stress state are needed. This corresponds to the solution of materials with linear elastic behavior. Therefore, using equations of equilibrium, stress-strain and strain-displacement, a differential equation for displacement is obtained and then the stresses at a time equal to zero are calculated. Using Norton’s law in the multi-axial form in conjunction with the above-mentioned equations in the rate form, the radial displacement rate is obtained and then the radial, circumferential and axial creep stress rates are calculated. When the stress rates are known, the stresses at any time are calculated iteratively. The analytical solution is obtained for the conditions of plane strain and plane stress. The thermal loading is as follows: inner surface is exposed to a uniform heat flux, and the outer surface is exposed to an airstream. The heat conduction equation for the one-dimensional problem in polar coordinates is used to obtain temperature distribution in the cylinder. The pressure, inner radius and outer radius are considered constant. Material properties are considered as constant. Following this, profiles are plotted for the radial displacements, radial stress, circumferential stress and axial stress as a function of radial direction and time.

  16. Comparison of attenuation coefficients for VVER-440 and VVER-1000 pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Marek, M.; Rataj, J.; Vandlik, S. [Reactor Physics Dept., Research Centre Rez, Husinec 130, 25068 (Czech Republic)

    2011-07-01

    The paper summarizes the attenuation coefficient of the neutron fluence with E > 0.5 MeV through a reactor pressure vessel for vodo-vodyanoi energetichesky reactor (VVER) reactor types measured and/or calculated for mock-up experiments, as well as for operated nuclear power plant (NPP) units. The attenuation coefficient is possible to evaluate directly only by using the retro-dosimetry, based on a combination of the measured activities from the weld sample and concurrent ex-vessel measurement. The available neutron fluence attenuation coefficients (E > 0.5 MeV), calculated and measured at a mock-up experiment simulating the VVER-440-unit conditions, vary from 3.5 to 6.15. A similar situation is used for the calculations and mock-up experiment measurements for the VVER-1000 RPV, where the attenuation coefficient of the neutron fluence varies from 5.99 to 8.85. Because of the difference in calculations for the real units and the mock-up experiments, the necessity to design and perform calculation benchmarks both for VVER-440 and VVER-1000 would be meaningful if the calculation model is designed adequately to a given unit. (authors)

  17. Manufacturing Cost Analysis of Novel Steel/Concrete Composite Vessel for Stationary Storage of High-Pressure Hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Zhili [ORNL; Zhang, Wei [ORNL; Wang, Jy-An John [ORNL; Ren, Fei [ORNL

    2012-09-01

    A novel, low-cost, high-pressure, steel/concrete composite vessel (SCCV) technology for stationary storage of compressed gaseous hydrogen (CGH2) is currently under development at Oak Ridge National Laboratory (ORNL) sponsored by DOE s Fuel Cell Technologies (FCT) Program. The SCCV technology uses commodity materials including structural steels and concretes for achieving cost, durability and safety requirements. In particular, the hydrogen embrittlement of high-strength low-alloy steels, a major safety and durability issue for current industry-standard pressure vessel technology, is mitigated through the use of a unique layered steel shell structure. This report presents the cost analysis results of the novel SCCV technology. A high-fidelity cost analysis tool is developed, based on a detailed, bottom-up approach which takes into account the material and labor costs involved in each of the vessel manufacturing steps. A thorough cost study is performed to understand the SCCV cost as a function of the key vessel design parameters, including hydrogen pressure, vessel dimensions, and load-carrying ratio. The major conclusions include: The SCCV technology can meet the technical/cost targets set forth by DOE s FCT Program for FY2015 and FY2020 for all three pressure levels (i.e., 160, 430 and 860 bar) relevant to the hydrogen production and delivery infrastructure. Further vessel cost reduction can benefit from the development of advanced vessel fabrication technologies such as the highly automated friction stir welding (FSW). The ORNL-patented multi-layer, multi-pass FSW can not only reduce the amount of labor needed for assembling and welding the layered steel vessel, but also make it possible to use even higher strength steels for further cost reductions and improvement of vessel structural integrity. It is noted the cost analysis results demonstrate the significant cost advantage attainable by the SCCV technology for different pressure levels when compared to the

  18. Pressure vessels design methods using the codes, fracture mechanics and multiaxial fatigue

    Directory of Open Access Journals (Sweden)

    Fatima Majid

    2016-10-01

    Full Text Available This paper gives a highlight about pressure vessel (PV methods of design to initiate new engineers and new researchers to understand the basics and to have a summary about the knowhow of PV design. This understanding will contribute to enhance their knowledge in the selection of the appropriate method. There are several types of tanks distinguished by the operating pressure, temperature and the safety system to predict. The selection of one or the other of these tanks depends on environmental regulations, the geographic location and the used materials. The design theory of PVs is very detailed in various codes and standards API, such as ASME, CODAP ... as well as the standards of material selection such as EN 10025 or EN 10028. While designing a PV, we must design the fatigue of its material through the different methods and theories, we can find in the literature, and specific codes. In this work, a focus on the fatigue lifetime calculation through fracture mechanics theory and the different methods found in the ASME VIII DIV 2, the API 579-1 and EN 13445-3, Annex B, will be detailed by giving a comparison between these methods. In many articles in the literature the uniaxial fatigue has been very detailed. Meanwhile, the multiaxial effect has not been considered as it must be. In this paper we will lead a discussion about the biaxial fatigue due to cyclic pressure in thick-walled PV. Besides, an overview of multiaxial fatigue in PVs is detailed

  19. Statistical analysis using the Bayesian nonparametric method for irradiation embrittlement of reactor pressure vessels

    Science.gov (United States)

    Takamizawa, Hisashi; Itoh, Hiroto; Nishiyama, Yutaka

    2016-10-01

    In order to understand neutron irradiation embrittlement in high fluence regions, statistical analysis using the Bayesian nonparametric (BNP) method was performed for the Japanese surveillance and material test reactor irradiation database. The BNP method is essentially expressed as an infinite summation of normal distributions, with input data being subdivided into clusters with identical statistical parameters, such as mean and standard deviation, for each cluster to estimate shifts in ductile-to-brittle transition temperature (DBTT). The clusters typically depend on chemical compositions, irradiation conditions, and the irradiation embrittlement. Specific variables contributing to the irradiation embrittlement include the content of Cu, Ni, P, Si, and Mn in the pressure vessel steels, neutron flux, neutron fluence, and irradiation temperatures. It was found that the measured shifts of DBTT correlated well with the calculated ones. Data associated with the same materials were subdivided into the same clusters even if neutron fluences were increased.

  20. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Science.gov (United States)

    Boåsen, Magnus; Efsing, Pål; Ehrnstén, Ulla

    2017-02-01

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects-the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations.

  1. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel

    Science.gov (United States)

    Monnet, Ghiath; Domain, Christophe; Queyreau, Sylvain; Naamane, Sanae; Devincre, Benoit

    2009-11-01

    The collective behavior of dislocations in reactor pressure vessel (RPV) steel involves dislocation properties on different phenomenological scales. In the multiscale approach, adopted in this work, we use atomic simulations to provide input data for larger scale simulations. We show in this paper how first-principles calculations can be used to describe the Peierls potential of screw dislocations, allowing for the validation of the empirical interatomic potential used in molecular dynamics simulations. The latter are used to compute the velocity of dislocations as a function of the applied stress and the temperature. The mobility laws obtained in this way are employed in dislocation dynamics simulations in order to predict properties of plastic flow, namely dislocation-dislocation interactions and dislocation interactions with carbides at low and high temperature.

  2. United States Department of Energy projects related to reactor pressure vessel annealing optimization

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T.; Nakos, J.T.

    1993-09-01

    Light water reactor pressure vessel (RPV) material properties reduced by long-term exposure to neutron irradiation can be recovered through a thermal annealing treatment. This technique to extend RPV life, discussed in this report, provides a complementary approach to analytical methodologies to evaluate RPV integrity. RPV annealing has been successfully demonstrated in the former Soviet Union and on a limited basis by the US (military applications only). The process of demonstrating the technical feasibility of annealing commercial US RPVs is being pursued through a cooperative effort between the nuclear industry and the US Department of Energy (USDOE) Plant Lifetime Improvement (PLIM) Program. Presently, two projects are under way through the USDOE PLIM Program to demonstrate the technical feasibility of annealing commercial US RPVS, (1) annealing re-embrittlement data base development and (2) heat transfer boundary condition experiments.

  3. Development of an ultrasonic imaging system for the inspection of nuclear reactor pressure vessels. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Becker, F.L.; Crow, V.L.; Davis, T.J.; Doctor, S.R.; Hildebrand, B.P.; Lemon, D.K.; Posakony, G.J.

    1979-10-01

    The development of an experimental model of an ultrasonic linear array system for the inspection of weldments in nuclear reactor pressure vessels is described. The imaging system is designed to operate in both pulse echo and holographic modes of operation. The system utilizes a sequentially pulsed, phase steered linear array to develop pulse echo images and a line focused illumination transducer in conjunction with a linear receiver array to develop holographic reconstructed images. The results recorded from the computer-based system demonstrate the capability of array technology. Excellent results from both the pulse echo and holographic modes of operation have been achieved. Pulse echo images of flaws in weldments are displayed in B-scan, C-scan, or isometric presentations. Reconstruction of the phase or holographic images are compared with pulse echo results and demonstrate the enhancement potential for the holographic procedure.

  4. Characterisation of creep cavitation damage in a stainless steel pressure vessel using small angle neutron scattering

    CERN Document Server

    Bouchard, P J; Treimer, W

    2002-01-01

    Grain-boundary cavitation is the dominant failure mode associated with initiation of reheat cracking, which has been widely observed in austenitic stainless steel pressure vessels operating at temperatures within the creep range (>450 C). Small angle neutron scattering (SANS) experiments at the LLB PAXE instrument (Saclay) and the V12 double-crystal diffractometer of the HMI-BENSC facility (Berlin) are used to characterise cavitation damage (in the size range R=10-2000 nm) in a variety of creep specimens extracted from ex-service plant. Factors that affect the evolution of cavities and the cavity-size distribution are discussed. The results demonstrate that SANS techniques have the potential to quantify the development of creep damage in type-316H stainless steel, and thereby link microstructural damage with ductility-exhaustion models of reheat cracking. (orig.)

  5. Qualification of non-destructive examination for belgian nuclear reactor pressure vessel inspection

    Energy Technology Data Exchange (ETDEWEB)

    Couplet, D. [TRACTEBEL, Brussels (Belgium); Francoise, T. [Intercontrole, 94 - Rungis (France)

    2001-07-01

    In Service Inspection (ISI) participates to the assessment of Nuclear Reactor Pressure Vessel Integrity. The performance of Non Destructive Examination (NDE) techniques must be demonstrated according to predefined objectives. The qualification process is essential to trust the reliability of the information provided by NDE. In Belgian Nuclear Power Plants, the qualification was conducted through a collaboration between the vendor and a technical group from the Electricity Utility. The important facts of this qualification will be presented: - the detailed definition of the inspection and qualifications objectives, based on a combination of the ASME code and the European Methodology for Qualification; - the systematic verification of the NDE performance and limitations, for each ISI objective, through an adequate combination of tests on blocks and technical justification; - the continuous improvement of the NDE procedure; - the feedback and the lessons learnt from site experience; - the necessary multi-disciplinary approach (NDE, degradation mechanisms, structural integrity)

  6. Microstructural investigations on Russian reactor pressure vessel steels by small-angle neutron scattering

    Science.gov (United States)

    Ulbricht, A.; Boehmert, J.; Strunz, P.; Dewhurst, C.; Mathon, M.-H.

    The effect of radiation embrittlement has a high safety significance for Russian VVER reactor pressure vessel steels. Heats of base and weld metals of the as-received state, irradiated state and post-irradiation annealed state were investigated using small-angle neutron scattering (SANS) to obtain insight about the microstructural features caused by fast neutron irradiation. The SANS intensities increase in the momentum transfer range between 0.8 and 3 nm-1 for all the material compositions in the irradiated state. The size distribution function of the irradiation-induced defect clusters has a pronounced maximum at 1 nm in radius. Their content varies between 0.1 and 0.7 vol.% dependent on material composition and increases with the neutron fluence. The comparison of nuclear and magnetic scattering indicates that the defects differ in their composition. Thermal annealing reduces the volume fraction of irradiation defect clusters.

  7. Repair weld induced residual stresses in thick-walled steel pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Smith, G.C.; Holz, P.P.

    1978-06-01

    If a flaw requiring corrective action were to be found in an operating nuclear pressure vessel, there would be considerable safety and economic implications. Should such a flaw be found, one possible corrective action would be an in situ repair weld. A repair of this type would presumably involve grinding away material in a region encompassing the flaw and then filling the resulting cavity with weld metal. Thermal stress relieving under those conditions could lead to serious difficulties associated with thermal expansion and warpage and would therefore most likely be avoided. Such a departure from normal procedure raises questions relating to residual stresses and material toughness levels which would have to be assessed before a repair could be recommended or approved. The residual stress measurements reported are intended to provide baseline information to aid in an assessment should such a repair ever have to be seriously considered.

  8. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-12-31

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program.

  9. Positron annihilation study of Fe-ion irradiated reactor pressure vessel model alloys

    Science.gov (United States)

    Chen, L.; Li, Z. C.; Schut, H.; Sekimura, N.

    2016-01-01

    The degradation of reactor pressure vessel steels under irradiation, which results from the hardening and embrittlement caused by a high number density of nanometer scale damage, is of increasingly crucial concern for safe nuclear power plant operation and possible reactor lifetime prolongation. In this paper, the radiation damage in model alloys with increasing chemical complexity (Fe, Fe-Cu, Fe-Cu-Si, Fe-Cu-Ni and Fe-Cu-Ni-Mn) has been studied by Positron Annihilation Doppler Broadening spectroscopy after 1.5 MeV Fe-ion implantation at room temperature or high temperature (290 oC). It is found that the room temperature irradiation generally leads to the formation of vacancy-type defects in the Fe matrix. The high temperature irradiation exhibits an additional annealing effect for the radiation damage. Besides the Cu-rich clusters observed by the positron probe, the results show formation of vacancy-Mn complexes for implantation at low temperatures.

  10. Stress Corrosion Cracking and Fatigue Crack Growth Studies Pertinent to Spacecraft and Booster Pressure Vessels

    Science.gov (United States)

    Hall, L. R.; Finger, R. W.

    1972-01-01

    This experimental program was divided into two parts. The first part evaluated stress corrosion cracking in 2219-T87 aluminum and 5Al-2.5Sn (ELI) titanium alloy plate and weld metal. Both uniform height double cantilever beam and surface flawed specimens were tested in environments normally encountered during the fabrication and operation of pressure vessels in spacecraft and booster systems. The second part studied compatibility of material-environment combinations suitable for high energy upper stage propulsion systems. Surface flawed specimens having thicknesses representative of minimum gage fuel and oxidizer tanks were tested. Titanium alloys 5Al-2.5Sn (ELI), 6Al-4V annealed, and 6Al-4V STA were tested in both liquid and gaseous methane. Aluminum alloy 2219 in the T87 and T6E46 condition was tested in fluorine, a fluorine-oxygen mixture, and methane. Results were evaluated using modified linear elastic fracture mechanics parameters.

  11. Application of advanced master curve approaches on WWER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner [Forschungszentrum Rossendorf e.V. (Germany)]. E-mail: h.w.viehrig@fz-rossendorf.de; Scibetta, Marc [SCK-CEN, Reactor Materials Research (Belgium); Wallin, Kim [VTT Industrial Systems, Materials and Structural Integrity (Finland)

    2006-08-15

    The master curve (MC) approach used to measure the transition temperature, T , was standarised in the ASTM Standard Test Method E 1921 in 1997. The basic MC approach for analysis of fracture test results is intended for macroscopically homogeneous steels with a body centred cubic (ferritic) structure only. In reality, due to the manufacturing process, the steels in question are seldom fully macroscopically homogeneous. The fracture toughness values measured on Charpy size SE(B) specimens of base metal from the Greifswald Unit 8 rector pressure vessel (RPV) show large scatter. The basic MC evaluation following ASTM E1921 supplies a MC with many fracture toughness values which lie below the 5% fracture probability line. It is therefore suspected that this material is macroscopically inhomogeneous. In this paper, two recent extensions of the MC for inhomogeneous materials are applied to these fracture toughness data.

  12. Design of pressure vessels. Part 2; Conception des enceintes sous pression. Partie 2

    Energy Technology Data Exchange (ETDEWEB)

    Grandemange, J.M. [Areva NP, 92 - Paris la Defense (France)

    2008-01-15

    This document deals with the classification of stresses, necessary for the implementation of the mechanical code criteria defined for the pressure vessels of PWR-type reactors. It describes the general approach of design, analysis, and in-service monitoring, the regulatory tests and the modalities of equivalence between industrial construction codes. Content: 1 - damage modes and stresses classification: context, general approach, example of application; 2 - from the design stage to the in-service monitoring: liabilities, design conditions, materials choice and dimensioning, analysis, particular case of pipes and valve parts, in-service monitoring; 3 - regulatory tests: context, tests prescribed by the design and construction rules of PWR mechanical components (RCC-M); 4 - equivalence possibilities between codes: codes for nuclear reactor equipments, convergence between industrial codes and standards; 5 - conclusion. (J.S.)

  13. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    Science.gov (United States)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  14. Atomic and dislocation dynamics simulations of plastic deformation in reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Monnet, Ghiath, E-mail: ghiathmonnet@yahoo.f [EDF-R and D, MMC, Avenue des Renardieres, 77818 Moret sur Loing (France); Domain, Christophe; Queyreau, Sylvain; Naamane, Sanae [EDF-R and D, MMC, Avenue des Renardieres, 77818 Moret sur Loing (France); Devincre, Benoit [LEM, CNRS-ONERA, 29 av. de la division Leclerc, 92130 Chatillon (France)

    2009-11-15

    The collective behavior of dislocations in reactor pressure vessel (RPV) steel involves dislocation properties on different phenomenological scales. In the multiscale approach, adopted in this work, we use atomic simulations to provide input data for larger scale simulations. We show in this paper how first-principles calculations can be used to describe the Peierls potential of screw dislocations, allowing for the validation of the empirical interatomic potential used in molecular dynamics simulations. The latter are used to compute the velocity of dislocations as a function of the applied stress and the temperature. The mobility laws obtained in this way are employed in dislocation dynamics simulations in order to predict properties of plastic flow, namely dislocation-dislocation interactions and dislocation interactions with carbides at low and high temperature.

  15. Pressure Vessel Fluence Calculations for the Hungarian VVER-440 Units for the Power Uprate and the Llifetime Extension

    Directory of Open Access Journals (Sweden)

    Hordósy Gábor

    2016-01-01

    Full Text Available A major project was launched at Paks NPP, Hungary, to investigate the possibility of lifetime extension up to 60 years. At the same time, new fuel types with higher enrichment and containing pins with gadolinium have been introduced. Due to these plans, the radiation load of the pressure vessel was evaluated up to 60 years irradiation, taking into account the past and planned future cycles. The computational procedure, elaborated and validated earlier for the fast flux calculation in the pressure vessel was modified for the new fuel types. The neutron source at the core boundaries was taken from core design calculations and the neutron transport from the source to and through the pressure vessel was followed by Monte Carlo calculations. A number of calculations were performed to adequately follow the change of the neutron source. The paper details this procedure, the used Monte Carlo model, the influence of the different reloading schemes on the radiation load and the calculated results.

  16. Pressure Vessel Fluence Calculations for the Hungarian VVER-440 Units for the Power Uprate and the Llifetime Extension

    Science.gov (United States)

    Hordósy, Gábor; Hegyi, György; Keresztúri, András; Maráczy, Csaba; Temesvári, Emese; Zsolnay, Éva M.

    2016-02-01

    A major project was launched at Paks NPP, Hungary, to investigate the possibility of lifetime extension up to 60 years. At the same time, new fuel types with higher enrichment and containing pins with gadolinium have been introduced. Due to these plans, the radiation load of the pressure vessel was evaluated up to 60 years irradiation, taking into account the past and planned future cycles. The computational procedure, elaborated and validated earlier for the fast flux calculation in the pressure vessel was modified for the new fuel types. The neutron source at the core boundaries was taken from core design calculations and the neutron transport from the source to and through the pressure vessel was followed by Monte Carlo calculations. A number of calculations were performed to adequately follow the change of the neutron source. The paper details this procedure, the used Monte Carlo model, the influence of the different reloading schemes on the radiation load and the calculated results.

  17. Spin Forming Aluminum Crew Module (CM) Metallic Aft Pressure Vessel Bulkhead (APVBH) - Phase II

    Science.gov (United States)

    Hoffman, Eric K.; Domack, Marcia S.; Torres, Pablo D.; McGill, Preston B.; Tayon, Wesley A.; Bennett, Jay E.; Murphy, Joseph T.

    2015-01-01

    The principal focus of this project was to assist the Multi-Purpose Crew Vehicle (MPCV) Program in developing a spin forming fabrication process for manufacture of the Orion crew module (CM) aft pressure vessel bulkhead. The spin forming process will enable a single piece aluminum (Al) alloy 2219 aft bulkhead resulting in the elimination of the current multiple piece welded construction, simplify CM fabrication, and lead to an enhanced design. Phase I (NASA TM-2014-218163 (1)) of this assessment explored spin forming the single-piece CM forward pressure vessel bulkhead. The Orion MPCV Program and Lockheed Martin (LM) recently made two critical decisions relative to the NESC Phase I work scope: (1) LM selected the spin forming process to manufacture a single-piece aft bulkhead for the Orion CM, and (2) the aft bulkhead will be manufactured from Al 2219. Based on the Program's new emphasis related to the spin forming process, the NESC was asked to conduct a Phase II assessment to assist in the LM manufacture of the aft bulkhead and to conduct a feasibility study into spin forming the Orion CM cone. This activity was approved on June 19, 2013. Dr. Robert Piascik, NASA Technical Fellow for Materials at the Langley Research Center (LaRC), was selected to lead this assessment. The project plan was approved by the NASA Engineering and Safety Center (NESC) Review Board (NRB) on July 18, 2013. The primary stakeholders for this assessment were the NASA and LM MPCV Program offices. Additional benefactors are commercial launch providers developing CM concepts.

  18. Rısk analysis for pressure vessel with external corrosion using RBI method based on API 581

    Science.gov (United States)

    Naubnome, Viktor; Haryadi, Gunawan Dwi; Ismail, Rifky; Kim, Seon Jin

    2016-04-01

    Internal corrosion and external are the one major cause of accidents in liquid and natural gas in a pressure vessel. To lessen the vessel risk level, many companies have adopted and applied risk based inspection (RBI) methodology to risk reduction equipment, This study applied RBI methodology to optimize the inspection planing of the pressure vessel in power plant unit Jawa-Bali. In API 581, the risk situation for each type of equipment was classified into four levels: low risk level, medium-risk level, medium-high-risk level, and high level. This is expressed as a risk matrix. In this paper, semi-quantitative analysis method of risk-based inspection (RBI) was carried out for reducing the failure level of risk and optimized inspection plans, risk analysis of equipment failures resulting from corrosion need to be implemented. The result RBI analysis showed that pressure vessel has a medium high risk level and medium level. Failure mechanisms that occur in the pressure vessel is general thinning.

  19. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  20. Common Defects of Pressure Vessel Welding and Prevention Measures%压力容器焊接常见缺陷的产生和防治措施

    Institute of Scientific and Technical Information of China (English)

    赵振芳

    2012-01-01

    The welding quality of pressure vessels is to ensure the safe operation of a pressure vessel key.The paper introduced several kinds of common pressure vessel welding defects and preventive measures.%压力容器焊接质量是保证压力容器安全运行的关键。文章介绍了压力容器焊接中几种常见缺陷并提出预防措施。

  1. Studies on the tempo of bubble formation in recently cavitated vessels: a model to predict the pressure of air bubbles.

    Science.gov (United States)

    Wang, Yujie; Pan, Ruihua; Tyree, Melvin T

    2015-06-01

    A cavitation event in a vessel replaces water with a mixture of water vapor and air. A quantitative theory is presented to argue that the tempo of filling of vessels with air has two phases: a fast process that extracts air from stem tissue adjacent to the cavitated vessels (less than 10 s) and a slow phase that extracts air from the atmosphere outside the stem (more than 10 h). A model was designed to estimate how water tension (T) near recently cavitated vessels causes bubbles in embolized vessels to expand or contract as T increases or decreases, respectively. The model also predicts that the hydraulic conductivity of a stem will increase as bubbles collapse. The pressure of air bubbles trapped in vessels of a stem can be predicted from the model based on fitting curves of hydraulic conductivity versus T. The model was validated using data from six stem segments each of Acer mono and the clonal hybrid Populus 84 K (Populus alba × Populus glandulosa). The model was fitted to results with root mean square error less than 3%. The model provided new insight into the study of embolism formation in stem tissue and helped quantify the bubble pressure immediately after the fast process referred to above. © 2015 American Society of Plant Biologists. All Rights Reserved.

  2. Instability and "Sausage-String" Appearance in Blood Vessels during High Blood Pressure

    CERN Document Server

    Alstrøm, P; Colding-Jorgensen, M; Gustafsson, F; Holstein-Rathlou, N H; Alstrom, Preben; Eguiluz, Victor M.; Colding-Jorgensen, Morten; Gustafsson, Finn; Holstein-Rathlou, Niels-Henrik

    1999-01-01

    A new Rayleigh-type instability is proposed to explain the `sausage-string' pattern of alternating constrictions and dilatations formed in blood vessels under influence of a vasoconstricting agent. Our theory involves the nonlinear elasticity characteristics of the vessel wall, and provides predictions for the conditions under which the cylindrical form of a blood vessel becomes unstable.

  3. Association of body composition and blood pressure categories with retinal vessel diameters in primary school children.

    Science.gov (United States)

    Imhof, Katharina; Zahner, Lukas; Schmidt-Trucksäss, Arno; Hanssen, Henner

    2016-06-01

    Alterations in retinal vessel diameters have been shown to be predictive of cardiovascular risk in adults and children. The aim of our study was to examine the association of body composition and blood pressure (BP) categories with retinal vessel diameters in school children. We examined anthropometric parameters, BP and retinal arteriolar (CRAE) and venular (CRVE) diameters as well as the arteriolar-to-venular diameter ratio (AVR) in 391 children (age: 7.3, s.d. 0.4). Differences between the lowest and highest BP quartiles indicated that higher systolic and diastolic BP were associated with narrower CRAE (P<0.001 for both). Children in the highest weight quartile had narrower CRAE compared with the lowest quartile (P=0.05). In the regression analysis, systolic and diastolic BP were associated with arteriolar narrowing (-0.4 measuring units (mu) per mm Hg, 95% confidence interval: [-0.6; -0.3] and -0.6 mu per mm Hg [-0.7; -0.4], respectively; P<0.001 for both). An independent association was found for diastolic BP only. Compared with normotensives (NT; 74.4% of cohort), arteriolar narrowing was already seen in children categorized as pre-hypertensive (PHT) (11.5% of cohort), which was similar to HT children (14.1% of cohort) (NT: mean 207.2 [205.6; 208.7] mu; PHT: 201.7 [197.8; 205.7] mu; HT: 199.7 [196.2; 203.3] mu; P=0.01 for PHT vs. NT and P<0.001 for HT vs. NT in systolic BP). Our results suggest that systolic and diastolic BP are main determinants of retinal arteriolar diameters; and therefore, microvascular health in young children. Pre-hypertension seems to be associated with retinal microvascular alterations early in life.

  4. Development of automated welding process for field fabrication of thick walled pressure vessels. Fourth quarter, FY 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-19

    Progress is reported in research on the automated welding of heavy steel plate for the fabrication of pressure vessels. Information is included on: torch and shield adaptation; mechanical control of the welding process; welding parameters; joint design; filler wire optimizaton; nondestructive testing of welds; and weld repair. (LCL)

  5. J-integral elastic plastic fracture mechanics evaluation of the stability of cracks in nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, M. P.; McMeeking, R. M.; Parks, D. M.

    1980-06-01

    Contributions were made toward developing a new methodology to assess the stability of cracks in pressure vessels made from materials that exhibit a significant increase in toughness during the early increments of crack growth. It has a wide range of validity from linear elastic to fully plastic behavior.

  6. Improvement in acupoint selection for acupuncture of nerves surrounding the injury site: electro-acupuncture with Governor vessel with local meridian acupoints

    Directory of Open Access Journals (Sweden)

    Guan-heng He

    2015-01-01

    Full Text Available Peripheral nerve injury not only affects the site of the injury, but can also induce neuronal apoptosis at the spinal cord. However, many acupuncture clinicians still focus only on the injury site, selecting acupoints entirely along the injured nerve trunk and neglecting other regions; this may delay onset of treatment efficacy and rehabilitation. Therefore, in the present study, we compared the clinical efficacy of acupuncture at Governor vessel and local meridian acupoints combined (GV/LM group with acupuncture at local meridian acupoints alone (LM group in the treatment of patients with peripheral nerve injury. In the GV/LM group (n = 15, in addition to meridian acupoints at the injury site, the following acupoints on the Governor vessel were stimulated: Baihui (GV20, Fengfu (GV16, Dazhui (GV14, and Shenzhu (GV12, selected to treat nerve injury of the upper limb, and Jizhong (GV6, Mingmen (GV4, Yaoyangguan (GV3, and Yaoshu (GV2 to treat nerve injury of the lower limb. In the LM group (n = 15, only meridian acupoints along the injured nerve were selected. Both groups had electroacupuncture treatment for 30 minutes, once a day, 5 times per week, for 6 weeks. Two cases dropped out of the LM group. A good or excellent clinical response was obtained in 80% of the patients in the GV/LM group and 38.5% of the LM group. In a second study, an additional 20 patients underwent acupuncture with the same prescription as the GV/LM group. Electomyographic nerve conduction tests were performed before and after acupuncture to explore the mechanism of action of the treatment. An effective response was observed in 80.0% of the patients, with greater motor nerve conduction velocity and amplitude after treatment, indicating that electroacupuncture on specific Governor vessel acupoints promotes functional motor nerve repair after peripheral nerve injury. In addition, electromyography was performed before, during and after electroacupuncture in one patient with

  7. Improvement in acupoint selection for acupuncture of nerves surrounding the injury site:electro-acupuncture with Governor vessel with local meridian acupoints

    Institute of Scientific and Technical Information of China (English)

    Guan-heng He; Jing-wen Ruan; Yuan-shan Zeng; Xin Zhou; Ying Ding; Guang-hui Zhou

    2015-01-01

    Peripheral nerve injury not only affects the site of the injury, but can also induce neuronal apop-tosis at the spinal cord. However, many acupuncture clinicians still focus only on the injury site, selecting acupoints entirely along the injured nerve trunk and neglecting other regions;this may delay onset of treatment efifcacy and rehabilitation. Therefore, in the present study, we compared the clinical efifcacy of acupuncture at Governor vessel and local meridian acupoints combined (GV/LM group) with acupuncture at local meridian acupoints alone (LM group) in the treatment of patients with peripheral nerve injury. In the GV/LM group (n = 15), in addition to meridian acupoints at the injury site, the following acupoints on the Governor vessel were stimulated:Baihui (GV20),Fengfu (GV16),Dazhui(GV14), andShenzhu (GV12), selected to treat nerve injury of the upper limb, andJizhong (GV6),Mingmen (GV4),Yaoyangguan (GV3), andYaoshu (GV2) to treat nerve injury of the lower limb. In the LM group (n = 15), only me-ridian acupoints along the injured nerve were selected. Both groups had electroacupuncture treatment for 30 minutes, once a day, 5 times per week, for 6 weeks. Two cases dropped out of the LM group. A good or excellent clinical response was obtained in 80% of the patients in the GV/LM group and 38.5% of the LM group. In a second study, an additional 20 patients underwent acupuncture with the same prescription as the GV/LM group. Electomyographic nerve conduc-tion tests were performed before and after acupuncture to explore the mechanism of action of the treatment. An effective response was observed in 80.0% of the patients, with greater motor nerve conduction velocity and amplitude after treatment, indicating that electroacupuncture on speciifc Governor vessel acupoints promotes functional motor nerve repair after peripheral nerve injury. In addition, electromyography was performed before, during and after electroacu-puncture in one patient with radial nerve

  8. Standard practice for examination of Gas-Filled filament-wound composite pressure vessels using acoustic emission

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice provides guidelines for acoustic emission (AE) examination of filament-wound composite pressure vessels, for example, the type used for fuel tanks in vehicles which use natural gas fuel. 1.2 This practice requires pressurization to a level equal to or greater than what is encountered in normal use. The tanks' pressurization history must be known in order to use this practice. Pressurization medium may be gas or liquid. 1.3 This practice is limited to vessels designed for less than 690 bar [10,000 psi] maximum allowable working pressure and water volume less than 1 m3 or 1000 L [35.4 ft3]. 1.4 AE measurements are used to detect emission sources. Other nondestructive examination (NDE) methods may be used to gain additional insight into the emission source. Procedures for other NDE methods are beyond the scope of this practice. 1.5 This practice applies to examination of new and in-service filament-wound composite pressure vessels. 1.6 This practice applies to examinations conducted at amb...

  9. Application of Master Curve fracture toughness for reactor pressure vessel integrity assessment in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Server, William; Rosinski, Stan; Lott, Randy; Kim, Charles; Weakland, Dennis

    2002-08-01

    The Master Curve fracture toughness approach has been used in the USA for better defining the transition temperature fracture toughness of irradiated reactor pressure vessel (RPV) steels for end-of-life (EOL) and EOL extension (EOLE) time periods. The first application was for the Kewaunee plant in which the life-limiting material was a circumferential weld metal. Fracture toughness testing of this weld metal corresponding to EOL and beyond EOLE was used to reassess the PTS screening value, RT{sub PTS}, and to develop new operating pressure-temperature curves. The NRC has approved this application using a shift-based methodology and higher safety margins than those proposed by the utility and its contractors. Beaver Valley Unit 1, a First Energy nuclear plant, has performed similar fracture toughness testing, but none of the testing has been conducted at EOL or EOLE at this time. Therefore, extrapolation of the life-limiting plate data to higher fluences is necessary, and the projections will be checked in the next decade by Master Curve fracture toughness testing of all of the Beaver Valley Unit 1 beltline materials (three plates and three welds) at fluences near or greater than EOLE. A supplemental surveillance capsule has been installed in the sister plant, Beaver Valley Unit 2, which has the capability of achieving a higher lead factor while operating under essentially the same environment. The Beaver Valley Unit 1 evaluation has been submitted to the NRC. This paper reviews the shift-based approach taken for the Beaver Valley Unit 1 RPV and presents the use of the RT{sub T{sub 0}} methodology (which evolved out of the Master Curve testing and endorsed through two ASME Code Cases). The applied margin accounts for uncertainties in the various material parameters. Discussion of a direct measurement of RT{sub T{sub 0}} approach, as originally submitted for the Kewaunee case, is also presented.

  10. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A. [Oak Ridge National Lab., TN (United States)] [and others

    1997-02-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.

  11. Interactions between dislocations and irradiation-induced defects in light water reactor pressure vessel steels

    Science.gov (United States)

    Jumel, Stéphanie; Van Duysen, Jean-Claude; Ruste, Jacky; Domain, Christophe

    2005-11-01

    The REVE project (REactor for Virtual Experiments) is an international effort aimed at developing tools to simulate irradiation effects in light water reactors materials. In the framework of this project, a European team developed a first tool, called RPV-1 designed for reactor pressure vessel steels. This article is the third of a series dedicated to the presentation of the codes and models used to build RPV-1. It describes the simplified approach adopted to simulate the irradiation-induced hardening. This approach relies on a characterization of the interactions between a screw dislocation and irradiation-induced defects from molecular dynamics simulations. The pinning forces exerted by the defects on the dislocation were estimated from the obtained results and some hypotheses. In RPV-1, these forces are used as input parameters of a Foreman and Makin-type code, called DUPAIR, to simulate the irradiation-induced hardening at 20 °C. The relevance of the proposed approach was validated by the comparison with experimental results. However, this work has to be considered as an initial step to facilitate the development of a first tool to simulate irradiation effects. It can be improved by many ways (e.g. by use of dislocation dynamics code).

  12. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Science.gov (United States)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  13. Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

    Directory of Open Access Journals (Sweden)

    Junxiao Zheng

    2016-01-01

    Full Text Available Maintaining the structural integrity of the reactor pressure vessel (RPV is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV or E>0.1 MeV at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes in XY plane leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes in XY plane. Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

  14. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    McHenry, H.I.; Alers, G.A. [National Inst. of Standards and Technology, Boulder, CO (United States). Materials Reliability Div.

    1998-03-01

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs.

  15. Nondestructive Methods and Special Test Instrumentation Supporting NASA Composite Overwrapped Pressure Vessel Assessments

    Science.gov (United States)

    Saulsberry, Regor; Greene, Nathanael; Cameron, Ken; Madaras, Eric; Grimes-Ledesma, Lorie; Thesken, John; Phoenix, Leigh; Murthy, Pappu; Revilock, Duane

    2007-01-01

    Many aging composite overwrapped pressure vessels (COPVs), being used by the National Aeronautics and Space Administration (NASA) are currently under evaluation to better quantify their reliability and clarify their likelihood of failure due to stress rupture and age-dependent issues. As a result, some test and analysis programs have been successfully accomplished and other related programs are still in progress at the NASA Johnson Space Center (JSC) White Sands Test Facility (WSTF) and other NASA centers, with assistance from the commercial sector. To support this effort, a group of Nondestructive Evaluation (NDE) experts was assembled to provide NDE competence for pretest evaluation of test articles and for application of NDE technology to real-time testing. Techniques were required to provide assurance that the test article had adequate structural integrity and manufacturing consistency to be considered acceptable for testing and these techniques were successfully applied. Destructive testing is also being accomplished to better understand the physical and chemical property changes associated with progression toward "stress rupture" (SR) failure, and it is being associated with NDE response, so it can potentially be used to help with life prediction. Destructive work also includes the evaluation of residual stresses during dissection of the overwrap, laboratory evaluation of specimens extracted from the overwrap to evaluate physical property changes, and quantitative microscopy to inform the theoretical micromechanics.

  16. A Comparison of Various Stress Rupture Life Models for Orbiter Composite Pressure Vessels and Confidence Intervals

    Science.gov (United States)

    Grimes-Ledesma, Lorie; Murthy, Pappu, L. N.; Phoenix, S. Leigh; Glaser, Ronald

    2006-01-01

    In conjunction with a recent NASA Engineering and Safety Center (NESC) investigation of flight worthiness of Kevlar Ovenvrapped Composite Pressure Vessels (COPVs) on board the Orbiter, two stress rupture life prediction models were proposed independently by Phoenix and by Glaser. In this paper, the use of these models to determine the system reliability of 24 COPVs currently in service on board the Orbiter is discussed. The models are briefly described, compared to each other, and model parameters and parameter error are also reviewed to understand confidence in reliability estimation as well as the sensitivities of these parameters in influencing overall predicted reliability levels. Differences and similarities in the various models will be compared via stress rupture reliability curves (stress ratio vs. lifetime plots). Also outlined will be the differences in the underlying model premises, and predictive outcomes. Sources of error and sensitivities in the models will be examined and discussed based on sensitivity analysis and confidence interval determination. Confidence interval results and their implications will be discussed for the models by Phoenix and Glaser.

  17. Application of the Master Curve approach for the irradiation embrittlement evaluation of pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, H.W.; Boehmert, J. [Forschungszentrum Rossendorf e.V., Inst. fuer Sicherheitsforschung, Dresden (Germany)

    2003-09-01

    The master curve (MC) approach and the associated reference temperature, T{sub 0}, as defined in the test standard ASTM E1921, is rapidly moving from the research laboratory to application in integrity assessment of components and structures. T{sub 0} is the index temperature for the universal MC, which considers the toughness behaviour of a specific material. ''The Structural Integrity Assessment Procedures for European Industry'' (SINTAP) contain a MC extension for analysing the fracture behaviour of inhomogeneous ferritic steels. This paper presents the application of the MC approach to the T{sub 0} determination of different types of Russian WWER-type reactor pressure vessel (RPV) steels. In addition the SINTAP-MC approach was applied to determine an alternative reference temperature, T{sub R}. The influence of different microstructures and compositions within one type of RPV steel and the effect of irradiation with fast neutrons on T{sub 0} are experimentally evaluated. In general the MC based T{sub 0} is about 72 K below the Charpy V-notch transition temperature related to an impact energy of 48 J. The paper demonstrates the application of MC based T{sub 0} and T{sub R} as an alternative reference temperature for neutron embrittled RPV steels used in the RPV integrity assessment. (orig.)

  18. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner [Forschungszentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany)], E-mail: H.W.Viehrig@fzd.de; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany)

    2009-04-15

    The master curve (MC) approach as standardised in the ASTM Standard Test Method E1921 was applied to weld metal of the reactor pressure vessel (RPV) beltline welding seam of Greifswald unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The orientation of the specimens within the welding seam is TL and TS according to ASTM E399. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the master curve. Nearly all values lie within the fracture toughness curves for 2% and 98% fracture probability. There is a strong variation of the reference temperature T{sub 0} through the thickness of the welding seam, which can be explained by microstructural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TL and TS orientation in the welding seam have a differentiating and integrating behaviour, respectively.

  19. Blood Pressure and Penumbral Sustenance in Stroke from Large Vessel Occlusion

    Science.gov (United States)

    Regenhardt, Robert W.; Das, Alvin S.; Stapleton, Christopher J.; Chandra, Ronil V.; Rabinov, James D.; Patel, Aman B.; Hirsch, Joshua A.; Leslie-Mazwi, Thabele M.

    2017-01-01

    The global burden of stroke remains high, and of the various subtypes of stroke, large vessel occlusions (LVOs) account for the largest proportion of stroke-related death and disability. Several randomized controlled trials in 2015 changed the landscape of stroke care worldwide, with endovascular thrombectomy (ET) now the standard of care for all eligible patients. With the proven success of this therapy, there is a renewed focus on penumbral sustenance. In this review, we describe the ischemic penumbra, collateral circulation, autoregulation, and imaging assessment of the penumbra. Blood pressure goals in acute stroke remain controversial, and we review the current data and suggest an approach for induced hypertension in the acute treatment of patients with LVOs. Finally, in addition to reperfusion and enhanced perfusion, efforts focused on developing therapeutic targets that afford neuroprotection and augment neural repair will gain increasing importance. ET has revolutionized stroke care, and future emphasis will be placed on promoting penumbral sustenance, which will increase patient eligibility for this highly effective therapy and reduce overall stroke-related death and disability. PMID:28717354

  20. Effects of Microstructural Inhomogeneity on Charpy Impact Properties for Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Song, Jaemin; Kim, Min-Chul; Choi, Kwon-Jae; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Reactor pressure vessel (RPV) steels are fabricated by vacuum carbon deoxidation (VCD), and then heat treatment of quenching and tempering is conducted after forging. The through-the-thickness variation of microstructure in RPV can occur due to the cooling rate gradient during quenching and inhomogeneous deformation during forging process. The variation of microstructure in RPV affects the mechanical properties, and inhomogeneity in mechanical properties can occur. The evaluation of mechanical properties of RPV is conducted at thickness of 1/4T. In order to evaluate the safety of RPV more correctly, the research about the through-the-thickness variation of microstructure and mechanical properties in RPV is need. 1. The fine low bainite (LB) is the dominant phase at the inner-surface (0T), but coarse upper bainite (UB) is the dominant phase at the center (1/2T). This is because cooling rate gradient from surface to center occurs during quenching. 2. Inter-lath carbides act as fracture initiation site, and it reduces impact toughness. 3. The upper shelf energy is low and the reference temperatures are high at the 1/4T. Impact properties are poor at 1/4T because of the formation of coarse upper bainite structure and coarse inter-lath carbides.

  1. Effect of tempering temperature on the microstructure and mechanical properties of a reactor pressure vessel steel

    Science.gov (United States)

    Li, C. W.; Han, L. Z.; Luo, X. M.; Liu, Q. D.; Gu, J. F.

    2016-08-01

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe3C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo2C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained.

  2. A Comparison of Various Stress Rupture Life Models for Orbiter Composite Pressure Vessels and Confidence Intervals

    Science.gov (United States)

    Grimes-Ledesma, Lorie; Murthy, Pappu L. N.; Phoenix, S. Leigh; Glaser, Ronald

    2007-01-01

    In conjunction with a recent NASA Engineering and Safety Center (NESC) investigation of flight worthiness of Kevlar Overwrapped Composite Pressure Vessels (COPVs) on board the Orbiter, two stress rupture life prediction models were proposed independently by Phoenix and by Glaser. In this paper, the use of these models to determine the system reliability of 24 COPVs currently in service on board the Orbiter is discussed. The models are briefly described, compared to each other, and model parameters and parameter uncertainties are also reviewed to understand confidence in reliability estimation as well as the sensitivities of these parameters in influencing overall predicted reliability levels. Differences and similarities in the various models will be compared via stress rupture reliability curves (stress ratio vs. lifetime plots). Also outlined will be the differences in the underlying model premises, and predictive outcomes. Sources of error and sensitivities in the models will be examined and discussed based on sensitivity analysis and confidence interval determination. Confidence interval results and their implications will be discussed for the models by Phoenix and Glaser.

  3. Dual-pump CARS of Air in a Heated Pressure Vessel up to 55 Bar and 1300 K

    Science.gov (United States)

    Cantu, Luca; Gallo, Emanuela; Cutler, Andrew D.; Danehy, Paul M.

    2014-01-01

    Dual-pump Coherent anti-Stokes Raman scattering (CARS) measurements have been performed in a heated pressure vessel at NASA Langley Research Center. Each measurement, consisting of 500 single shot spectra, was recorded at a fixed location in dry air at various pressures and temperatures, in a range of 0.03-55×10(exp 5) Pa and 300-1373 K, where the temperature was varied using an electric heater. The maximum output power of the electric heater limited the combinations of pressures and temperatures that could be obtained. Charts of CARS signal versus temperature (at constant pressure) and signal versus pressure (at constant temperature) are presented and fit with an empirical model to validate the range of capability of the dual-pump CARS technique; averaged spectra at different conditions of pressure and temperature are also shown.

  4. Preliminary investigation of an ultrasound method for estimating pressure changes in deep-positioned vessels

    DEFF Research Database (Denmark)

    Olesen, Jacob Bjerring; Villagómez Hoyos, Carlos Armando; Traberg, Marie Sand

    2016-01-01

    varying from -40 Pa to 15 Pa with a standard deviation of 32%, and a bias of 25% found relative to the peak simulated pressure drop. This preliminary study shows that pressure can be estimated non-invasively to a depth that enables cardiac scans, and thereby, the possibility of detecting the pressure......This paper presents a method for measuring pressure changes in deep-tissue vessels using vector velocity ultrasound data. The large penetration depth is ensured by acquiring data using a low frequency phased array transducer. Vascular pressure changes are then calculated from 2-D angle...... of angle-independent vector velocities are acquired using directional synthetic aperture vector flow imaging. The obtained results are evaluated by comparison to a 3-D numerical simulation model with equivalent geometry as the designed phantom. The study showed pressure drops across the constricted phantom...

  5. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  6. Bobbin-Tool Friction-Stir Welding of Thick-Walled Aluminum Alloy Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Dalder, E C; Pastrnak, J W; Engel, J; Forrest, R S; Kokko, E; Ternan, K M; Waldron, D

    2007-06-06

    It was desired to assemble thick-walled Al alloy 2219 pressure vessels by bobbin-tool friction-stir welding. To develop the welding-process, mechanical-property, and fitness-for-service information to support this effort, extensive friction-stir welding-parameter studies were conducted on 2.5 cm. and 3.8 cm. thick 2219 Al alloy plate. Starting conditions of the plate were the fully-heat-treated (-T62) and in the annealed (-O) conditions. The former condition was chosen with the intent of using the welds in either the 'as welded' condition or after a simple low-temperature aging treatment. Since preliminary stress-analyses showed that stresses in and near the welds would probably exceed the yield-strength of both 'as welded' and welded and aged weld-joints, a post-weld solution-treatment, quenching, and aging treatment was also examined. Once a suitable set of welding and post-weld heat-treatment parameters was established, the project divided into two parts. The first part concentrated on developing the necessary process information to be able to make defect-free friction-stir welds in 3.8 cm. thick Al alloy 2219 in the form of circumferential welds that would join two hemispherical forgings with a 102 cm. inside diameter. This necessitated going to a bobbin-tool welding-technique to simplify the tooling needed to react the large forces generated in friction-stir welding. The bobbin-tool technique was demonstrated on both flat-plates and plates that were bent to the curvature of the actual vessel. An additional issue was termination of the weld, i.e. closing out the hole left at the end of the weld by withdrawal of the friction-stir welding tool. This was accomplished by friction-plug welding a slightly-oversized Al alloy 2219 plug into the termination-hole, followed by machining the plug flush with both the inside and outside surfaces of the vessel. The second part of the project involved demonstrating that the welds were fit for the intended

  7. D0 Silicon Upgrade: Gas Helium Storage Tank Pressure Vessel Engineering Note

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, Russ; /Fermilab

    1996-11-11

    This is to certify that Beaird Industries, Inc. has done a white metal blast per SSPC-SP5 as required per specifications on the vessel internal. Following the blast, a black light inspection was performed by Beaird Quality Control personnel to assure that all debris, grease, etc. was removed and interior was clean prior to closing vessel for helium test.

  8. Safety-related considerations for reactor pressure vessels in consideration of hydrogen flaking

    Energy Technology Data Exchange (ETDEWEB)

    Dugan, S.; Herter, K.H.; Schuler, X.; Silcher, H. [Stuttgart Univ. (Germany). MPA

    2013-07-01

    During non-destructive inspection of the reactor pressure vessels in the Belgian nuclear power plants Doel 3 and Tihange 2, a large number of crack-like indications located in the base metal of the core shells were found. As part of the evaluation of these indications, which were identified as flake-like separations (hydrogen flakes), questions arise as to their cause, possible operational growth and the impact on the continued safe operation of the plant. In addition to the operational load cases, possible accidental and beyond design load cases are also of importance. Within the scope of the ''Research Project Component Safety'' (Forschungsvorhaben Komponentensicherheit - FKS) in the time frame mid-1970s to mid-1990s, numerous R and D activities on the material mechanics behavior and qualification of RPV materials were performed at MPA University of Stuttgart. The objectives of these investigations were focused on material mechanical issues related to the integrity of components and included standard material testing as well as component-like large scale specimen tests. Another major objective was the evaluation of non-destructive testing (NDT) methods with respect to their detection capabilities for such defects which developed during the manufacturing process. The investigations also included a study of the conditions favorable for formation or prevention of hydrogen flaking. In the context of this paper, the results from these R and D activities are presented in view of the current issues and in relation to the integrity concept for German RPVs. Ultrasonic testing (UT) techniques applied during manufacturing and during in-service inspections of German RPVs will also be discussed.

  9. Closed vessel combustion modelling by using pressure-time evolution function derived from two-zonal approach

    Directory of Open Access Journals (Sweden)

    Tomić Mladen A.

    2012-01-01

    Full Text Available In this paper a new method for burned mass fraction - pressure relation, x-p relation, for two-zone model combustion calculation is developed. The main application of the two-zone model is obtaining laminar burning velocity, SL, by using a pressure history from a closed vessel combustion experiment. The linear x-p relation by Lewis and Von Elbe is still widely used. For linear x-p relation, the end pressure is necessary as input data for the description of the combustion process. In this paper a new x-p relation is presented on the basis of mass and energy conservation during the combustion. In order to correctly represent pressure evolution, the model proposed in this paper needs several input parameters. They were obtained from different sources, like the PREMIX software (with GRIMECH 3.0 mechanism and GASEQ software, as well as thermodynamic tables. The error analysis is presented in regard to the input parameters. The proposed model is validated against the experiment by Dahoe and Goey, and compared with linear x-p relation from Lewis and Von Elbe. The proposed two zone model shows sufficient accuracy when describing the combustion process in a closed vessel without knowing the end pressure in advance, i.e. both peak pressure and combustion rates can be sufficiently correctly captured.

  10. Low Background Stainless Steel for the Pressure Vessel in the PandaX-II Dark Matter Experiment

    CERN Document Server

    Zhang, Tao; Ji, Xiangdong; Liu, Jianglai; Liu, Xiang; Wang, Xuming; Yao, Chunfa; Yuan, Xunhua

    2016-01-01

    We report on the custom produced low radiation background stainless steel and the welding rod for the PandaX experiment, one of the deep underground experiments to search for dark matter and neutrinoless double beta decay using xenon. The anthropogenic 60 Co concentration in these samples is at the range of 1 mBq/kg or lower. We also discuss the radioactivity of nuclear-grade stainless steel from TISCO which has a similar background rate. The PandaX-II pressure vessel was thus fabricated using the stainless steel from CISRI and TISCO. Based on the analysis of the radioactivity data, we also made discussions on potential candidate for low background metal materials for future pressure vessel development.

  11. Views of TAGSI on the effects of gamma irradiation on the mechanical properties of irradiated ferritic steel reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Engineering, Metallurgy and Materials, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); English, C.A. [Materials and Chemistry Consultancy, Nexia Solutions, 168 Harwell International Business Centre, Didcot, Oxon OX11 0QJ (United Kingdom); Weaver, D.R. [School of Physics, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Assurance, Walton House, 404 The Quadrant, Birchwood Park, Warrington, Cheshire WA3 6AT (United Kingdom)

    2005-12-01

    The paper reviews and analyses the effects of gamma irradiation dose on the properties of ferritic steels used in reactor pressure vessels (RPVs). It explains factors that affect the embrittlement of a RPV steel induced by combinations of fast neutrons, thermal neutrons, and gamma irradiation. TAGSI were asked to consider the effects of gamma irradiation dose on the properties of steels used in reactor pressure vessels. TAGSI endorsed the use of the MCBEND code to calculate gamma fluxes and energetic gamma ray displacement cross-sections calculated using either Baumann or Alexander methods. TAGSI endorsed the calculation of the materials property changes due to an additional gamma dose using trend curves based on empirical correlation to neutron-induced damage (where k {sub {gamma}}{approx}1{+-}0.25)

  12. Low background stainless steel for the pressure vessel in the PandaX-II dark matter experiment

    Science.gov (United States)

    Zhang, T.; Fu, C.; Ji, X.; Liu, J.; Liu, X.; Wang, X.; Yao, C.; Yuan, Xunhua

    2016-09-01

    We report on the custom produced low radiation background stainless steel and the welding rod for the PandaX experiment, one of the deep underground experiments to search for dark matter and neutrinoless double beta decay using xenon. The anthropogenic 60Co concentration in these samples is at the range of 1 mBq/kg or lower. We also discuss the radioactivity of nuclear-grade stainless steel from TISCO which has a similar background rate. The PandaX-II pressure vessel was thus fabricated using the stainless steel from CISRI and TISCO. Based on the analysis of the radioactivity data, we also made discussions on potential candidate for low background metal materials for future pressure vessel development.

  13. Exact and Numerical Elastic Analysis for the FGM Thick-Walled Cylindrical Pressure Vessels with Exponentially-Varying Properties

    Directory of Open Access Journals (Sweden)

    Nejad M. Zamani

    2016-09-01

    Full Text Available Assuming exponential-varying properties in the radial direction and based on the elasticity theory, an exact closed-form analytical solution is obtained to elastic analysis of FGM thick-walled cylindrical pressure vessels in the plane strain condition. Following this, radial distribution of radial displacement, radial stress, and circumferential stress are plotted for different values of material inhomogeneity constant. The displacements and stresses distributions are compared with the solutions of the finite element method (FEM.

  14. Mechanical properties of the as-forged and the forged-and-milled steels for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Bong Sang; Yoon, Ji Hyun; Kim, Joo Hak; Oh, Yong Jun; Hong, Jun Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-04-01

    The mechanical properties of the as-forged and the forged and milled SA508-Gr.3 reactor pressure vessel steels were evaluated. The full Charpy impact curves obtained for four different locations in test materials. The various data including yield strengths, tensile strengths, elongations were obtained from the tensile strengths, elongations were obtained from the tensile test results for two locations in test materials. The detailed test results were integrated and analysed in this report. 6 refs., 7 figs., 5 tabs. (Author)

  15. Blood pressure regulation V: in vivo mechanical properties of precapillary vessels as affected by long-term pressure loading and unloading.

    Science.gov (United States)

    Eiken, Ola; Mekjavic, Igor B; Kölegård, Roger

    2014-03-01

    Recent studies are reviewed, concerning the in vivo wall stiffness of arteries and arterioles in healthy humans, and how these properties adapt to iterative increments or sustained reductions in local intravascular pressure. A novel technique was used, by which arterial and arteriolar stiffness was determined as changes in arterial diameter and flow, respectively, during graded increments in distending pressure in the blood vessels of an arm or a leg. Pressure-induced increases in diameter and flow were smaller in the lower leg than in the arm, indicating greater stiffness in the arteries/arterioles of the leg. A 5-week period of intermittent intravascular pressure elevations in one arm reduced pressure distension and pressure-induced flow in the brachial artery by about 50%. Conversely, prolonged reduction of arterial/arteriolar pressure in the lower body by 5 weeks of sustained horizontal bedrest, induced threefold increases of the pressure-distension and pressure-flow responses in a tibial artery. Thus, the wall stiffness of arteries and arterioles are plastic properties that readily adapt to changes in the prevailing local intravascular pressure. The discussion concerns mechanisms underlying changes in local arterial/arteriolar stiffness as well as whether stiffness is altered by changes in myogenic tone and/or wall structure. As regards implications, regulation of local arterial/arteriolar stiffness may facilitate control of arterial pressure in erect posture and conditions of exaggerated intravascular pressure gradients. That increased intravascular pressure leads to increased arteriolar wall stiffness also supports the notion that local pressure loading may constitute a prime mover in the development of vascular changes in hypertension.

  16. Investigation of low-cycle fatigue behavior of austenitic stainless steel for cold-stretched pressure vessels

    Institute of Scientific and Technical Information of China (English)

    Cun-jian MIAO; Jin-yang ZHENG; Xiao-zhe GAO; Ze HUANG; A-bin GUO; Du-yi YE; Li MA

    2013-01-01

    Cold-stretched pressure vessels from austenitic stainless steels (ASS) are widely used for storage and transportation of liquefied gases,and have such advantages as thin wall and light weight.Fatigue is an important concern in these pressure vessels,which are subjected to alternative loads.Even though several codes and standards have guidelines on these pressure vessels,there are no relevant design methods on fatigue failure.To understand the fatigue properties of ASS 1.4301 (equivalents include UNS S30400 and AISI 304) in solution-annealed (SA) and cold-stretched conditions (9% strain level) and the response of fatigue properties to cold stretching (CS),low-cycle fatigue (LCF) tests were performed at room temperature,with total strain amplitudes ranging from ±0.4% to ±0.8%.Martensite transformations were measured during the tests.Comparisons on cyclic stress response,cyclic stress-strain behavior,and fatigue life were carried out between SA and CS materials.Results show that CS reduces the initial hardening stage,but prolongs the softening period in the cyclic stress response.Martensite transformation helps form a stable regime and subsequent secondary hardening.The stresses of monotonic and cyclic stress-strain curves are improved by CS,which leads to a lower plastic strain and a much higher elastic strain.The fatigue resistance of the CS material is better than that of the SA material,which is approximately 1 x l03 to 2×104 cycles.The S-N curve of the ASME standard for ASS is compared with the fatigue data and is justified to be suitable for the fatigue design of cold-stretched pressure vessels.However,considering the CS material has a better fatigue resistance,the S-N curve will be more conservative.The present study would be helpful in making full use of the advantages of CS to develop a new S-N curve for fatigue design of cold-stretched pressure vessels.

  17. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  18. 46 CFR 35.25-5 - Repairs of boilers and unfired pressure vessels and reports of repairs or accidents by chief...

    Science.gov (United States)

    2010-10-01

    ... reports of repairs or accidents by chief engineer-TB/ALL. 35.25-5 Section 35.25-5 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY TANK VESSELS OPERATIONS Engine Department § 35.25-5 Repairs of boilers and unfired pressure vessels and reports of repairs or accidents by chief engineer—TB/ALL. (a) Before...

  19. Fatigue test of carbon epoxy composite high pressure hydrogen storage vessel under hydrogen environment

    Institute of Scientific and Technical Information of China (English)

    Chuan-xiang ZHENG; Liang WANG; Rong LI; Zong-xin WEI; Wei-wei ZHOU

    2013-01-01

    A significant temperature raise within hydrogen vehicle cylinder during the fast filling process will be observed,while the strength and fatigue life of the cylinder will dramatically decrease at high temperature.In order to evaluate the strength and fatigue of composite hydrogen storage vessel,a 70-MPa fatigue test system using hydrogen medium was set up.Experimental study on the fatigue of composite hydrogen storage vessels under real hydrogen environment was performed.The experimental results show that the ultimate strength and fatigue life both decreased obviously compared with the values under hydraulic fatigue test.Furthermore,fatigue property,failure behavior,and safe hydrogen charging/discharging working mode of onboard hydrogen storage vessels were obtained through the fatigue tests.

  20. Development of automated welding process for field fabrication of thick walled pressure vessels. Fourth quarter technical progress report for period ending September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-01-01

    Progress in developing an automated welding process for the field fabrication of thick walled pressure vessels is reported. Plans for the demonstration facility, for nondestructive testing, and for the procurement of materials are discussed. (LCL)

  1. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... vessels. 57.13001 Section 57.13001 Mineral Resources MINE SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR METAL AND NONMETAL MINE SAFETY AND HEALTH SAFETY AND HEALTH STANDARDS-UNDERGROUND METAL AND... the standards and specifications of the American Society of Mechanical Engineers Boiler and...

  2. In vivo quantification of lymph viscosity and pressure in lymphatic vessels and draining lymph nodes of arthritic joints in mice.

    Science.gov (United States)

    Bouta, Echoe M; Wood, Ronald W; Brown, Edward B; Rahimi, Homaira; Ritchlin, Christopher T; Schwarz, Edward M

    2014-03-15

    Rheumatoid arthritis (RA) is a chronic inflammatory joint disease with episodic flares. In TNF-Tg mice, a model of inflammatory-erosive arthritis, the popliteal lymph node (PLN) enlarges during the pre-arthritic 'expanding' phase, and then 'collapses' with adjacent knee flare associated with the loss of the intrinsic lymphatic pulse. As the mechanisms responsible are unknown, we developed in vivo methods to quantify lymph viscosity and pressure in mice with wild-type (WT), expanding and collapsed PLN. While no differences in viscosity were detected via multiphoton fluorescence recovery after photobleaching (MP-FRAP) of injected FITC-BSA, a 32.6% decrease in lymph speed was observed in vessels afferent to collapsed PLN (P pressure (LNP) demonstrated a decrease in expanding PLN versus WT pressure (3.41 ± 0.43 vs. 6.86 ± 0.56 cmH2O; P pressure (LPP), measured indirectly by slowly releasing a pressurized cuff occluding indocyanine green (ICG), demonstrated an increase in vessels afferent to expanding PLN versus WT (18.76 ± 2.34 vs. 11.04 ± 1.47 cmH2O; P pressure, and provide evidence to support the hypothesis that lymphangiogenesis and lymphatic transport are compensatory mechanisms to prevent synovitis via increased drainage of inflamed joints. Furthermore, the decrease in lymphatic flow and loss of LPP during PLN collapse are consistent with decreased drainage from the joint during arthritic flare, and validate these biomarkers of RA progression and possibly other chronic inflammatory conditions.

  3. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1989-01-01

    Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.

  4. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  5. Influence of surrounding gas, composition and pressure on plasma plume dynamics of nanosecond pulsed laser-induced aluminum plasmas

    Directory of Open Access Journals (Sweden)

    Mahmoud S. Dawood

    2015-10-01

    Full Text Available In this article, we present a comprehensive study of the plume dynamics of plasmas generated by laser ablation of an aluminum target. The effect of both ambient gas composition (helium, nitrogen or argon and pressure (from ∼5 × 10−7 Torr up to atmosphere is studied. The time- and space- resolved observation of the plasma plume are performed from spectrally integrated images using an intensified Charge Coupled Device (iCCD camera. The iCCD images show that the ambient gas does not significantly influence the plume as long as the gas pressure is lower than 20 Torr and the time delay below 300 ns. However, for pressures higher than 20 Torr, the effect of the ambient gas becomes important, the shortest plasma plume length being observed when the gas mass species is highest. On the other hand, space- and time- resolved emission spectroscopy of aluminum ions at λ = 281.6 nm are used to determine the Time-Of-Flight (TOF profiles. The effect of the ambient gas on the TOF profiles and therefore on the propagation velocity of Al ions is discussed. A correlation between the plasma plume expansion velocity deduced from the iCCD images and that estimated from the TOF profiles is presented. The observed differences are attributed mainly to the different physical mechanisms governing the two diagnostic techniques.

  6. Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E. [Oak Ridge National Lab., Heavy-Section Steel Technology Program, Oak Ridge, TN (United States)

    2001-07-01

    A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

  7. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Potirniche, Gabriel [Univ. of Idaho, Moscow, ID (United States); Barlow, Fred D. [Univ. of Idaho, Moscow, ID (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Rink, Karl [Univ. of Idaho, Moscow, ID (United States)

    2013-11-26

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  8. Development of an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J.C.; Choi, J.B.; Kim, Y.J. [SAFE Research Center, Sungkyunkwan Univ., Kyunggi-do (Korea); Choi, Y.H.; Park, Y.W. [Korea Inst. of Nuclear Safety, Daijon (Korea); Yoshimura, S. [Inst. of Environmental Studies, The Univ. of Tokyo, Tokyo (Japan)

    2004-07-01

    Since early 1950's, the fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, and as a result, various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (information technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of locations. For this reason, a network system based on internet or intranet bas been appeared in various fields of business. Evaluating the integrity of critical components is one of the most critical issues in the nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including periodical in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adopts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses virtual reality (VR) technique, virtual network computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and provide experts to co-operate each other by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel. (orig.)

  9. Nitrogen-Bearing Stainless Steels for Pressure Vessel%压力容器用含氮不锈钢

    Institute of Scientific and Technical Information of China (English)

    黄嘉琥

    2013-01-01

    压力容器用不锈钢几乎全为奥氏体不锈钢和奥氏体/铁素体双相不锈钢。在ASME -2011a, EN 13445:2009及GB 150-2011压力容器标准中所采用的奥氏体与双相不锈钢的所有牌号中,含氮钢牌号所占百分比分别为63.4%,76.4%和60%。在不锈钢材料标准中所有的双相不锈钢(指1971年后研制的牌号)、超级奥氏体不锈钢以及超级或特超级双相不锈钢中的氮含量基本均为0.1%~0.6%(称其为中氮型不锈钢)。EN 13445:2009规定可用于-273℃的奥氏体不锈钢10个牌号均为含氮钢。本文讨论了氮在不锈钢中的溶解度与含量、含氮不锈钢的类型、氮对不锈钢性能与组织的影响以及含氮不锈钢在压力容器中的应用。国外压力容器标准中N=0.1%~0.6%的中氮型奥氏体不锈钢已成为最重要的高性能不锈钢。这些钢在中国压力容器标准中尚未采用。而ASME-2011a中已用12个牌号,EN 13445:2009中已用14个牌号。中国仅在GB/T 20878-2007中已有23个牌号,GB/T 4237-2007(或GB/T 3280-2007)中已有13个牌号。本文建议在此基础上可进行更多的压力容器的应用工作。%Types of stainless steel for pressure vessel almost are all austenitic stainless steel and duplex austenitic/ferritic stainless steel.The percentages of designations of nitrogen-bearing steel grade individ-ually are 63 .4%,76 .4% and 60% among all designations of austenitic and duplex stainless steel types used in ASME-2011a,EN 13445:2009 and GB 150-2011 pressure vessel standards.Nitrogen contents of all duplex stainless steels(research after 1971),all super austenitic stainless steels and all super or hy-per duplex stainless steels in stainless steel standards basically are 0.1%~0.6%(call them middle ni-trogen-bearing stainless steel).10 designations of austenitic stainless steel can used for -273 ℃speci-fied in EN 13445:2009 are all nitrogen-bearing steel grade

  10. Comparative assessment of cyclic J-R curve determination by different methods in a pressure vessel steel

    Science.gov (United States)

    Chowdhury, Tamshuk; Sivaprasad, S.; Bar, H. N.; Tarafder, S.; Bandyopadhyay, N. R.

    2016-04-01

    Cyclic J-R behaviour of a reactor pressure vessel steel using different methods available in literature has been examined to identify the best suitable method for cyclic fracture problems. Crack opening point was determined by moving average method. The η factor was experimentally determined for cyclic loading conditions and found to be similar to that of ASTM value. Analyses showed that adopting a procedure analogous to the ASTM standard for monotonic fracture is reasonable for cyclic fracture problems, and makes the comparison to monotonic fracture results straightforward.

  11. Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Erak, D. Yu.; Lavrenchuk, O. V.

    2002-02-01

    Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.

  12. Study on surface nanocrystallization and resisting H2S stresscorrosion properties of pressure vessel steel welding joints

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Many efforts were spent on the homogenization of microstructure and property of welding joints. A new surface nanocrystallization technique named Supersonic Particles Bombarding(SSPB) can be used for this purpose. Two kinds of pressure vessel steel welding joints, 16MnR and 0Cr18Ni9Ti, were chosen to be treated by SSPB. Transmission electron microscopy was introduced to examine the surface microstructure. And their ability to resist H2 S stress corrosion was enhanced significantly after the SSPB treatment. The mechanism for the results were analyzed as well.

  13. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  14. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Science.gov (United States)

    Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.

    2016-02-01

    Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  15. Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

    Directory of Open Access Journals (Sweden)

    Dzhalandinov A.

    2016-01-01

    Full Text Available Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

  16. Blood pressure and sodium: Association with MRI markers in cerebral small vessel disease

    OpenAIRE

    Heye, Anna K.; Thrippleton, Michael J; Chappell, Francesca M; Valdés Hernández, Maria del C.; Armitage, Paul A.; Makin, Stephen D.; Muñoz Maniega, Susana; Sakka, Eleni; Flatman, Peter W.; Dennis, Martin S.; Wardlaw, Joanna M.

    2016-01-01

    Dietary salt intake and hypertension are associated with increased risk of cardiovascular disease including stroke. We aimed to explore the influence of these factors, together with plasma sodium concentration, in cerebral small vessel disease (SVD). In all, 264 patients with nondisabling cortical or lacunar stroke were recruited. Patients were questioned about their salt intake and plasma sodium concentration was measured; brain tissue volume and white-matter hyperintensity (WMH) load were m...

  17. Development of Long Shank Repair Tool for Defect of Pressure Vessel Bolt Hole in Pressure Vessel of Nuclear Power Plant%核电站压力容器螺孔长杆梳刀装置研制

    Institute of Scientific and Technical Information of China (English)

    黄新东; 黄辉; 洪龙; 李鑫

    2013-01-01

    In the reactor operation and operations with open lid,various defects may emerge on thread section of the main bolt hole pressure vessel.These defects must be dealt with before the closing of the lid.In view of the above conditions,this paper developed a pressure vessel screw rod cutter device,and expounds the long shank repair tool,a detailed description of the design scheme of the tool as well as the concrete structure.%在反应堆运行和开盖操作过程中,压力容器的主螺栓孔的螺纹段可能会产生各种缺陷,这些缺陷在再次扣盖前必须经过处理.针对上述工况,研制了一种压力容器螺孔长杆梳刀装置,本文阐述了该长杆梳刀装置的设计要求,详细描述该装置的设计方案以及具体结构形式.

  18. A cylindrical multiwire high-pressure gas proportional chamber surrounding a gaseous $_{2} target with a mylar separation foil $6 \\mu m thick

    CERN Document Server

    Gastaldi, Ugo; Averdung, H; Bailey, J; Beer, G A; Dreher, B; Erdman, K L; Klempt, E; Merle, K; Neubecker, K; Sabev, C; Schwenk, H; Wendling, R D; White, B L; Wodrich, R

    1978-01-01

    The characteristics and performances of a cylindrical multiwire proportional chamber built and used at CERN in experiment S142 for the study of the pp atom spectroscopy are presented. The chamber surrounds a high-pressure gaseous H/sub 2/ target, from which it is separated by a very thin window (6 mu m mylar foil). The active volume (90 cm long; 2 cm thick, internal diameter=30 cm) is divided into 36 equal and independent cells each covering 10 degrees in azimuth. At 4 abs. atm the detection efficiency for X-rays is higher than 20% in the whole energy range 1.5-15 keV. Typical resolutions are 35% fwhm for the 3 ke V Ar fluorescence line and 25% fwhm for the 5.5 keV /sup 54/Mn line. Working pressures from 0.5 to 16 abs. atm have been used. (8 refs).

  19. Sialyte(TM)-Based Composite Pressure Vessels for Extreme Environments Project

    Data.gov (United States)

    National Aeronautics and Space Administration — While traveling to Venus, electronics and instruments go through enormous pressure, temperature, and atmospheric environment changes. In the past, this has caused...

  20. Status and Perspectives of Nuclear Reactor Pressure Vessel Life Extension up to 60 Years Operation in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Chaouadi, R.; Scibetta, M.; Van Walle, E.

    2009-09-15

    The scope of this report involves a safety evaluation of the reactor pressure vessel (RPV) against neutron embrittlement, in the most severely irradiation region (belt line) and in the event of a pressurized thermal shock. The irreplaceable RPV is considered to be the most critical component for lifetime considerations of the nuclear power plant. However, an application for operation extension will also depend upon a number of additional considerations, including the technical assessment of other plant components, as well as non-technical arguments (e.g. political, environmental, economical, strategical that are outside the scope this report. In the hypothesis of a request for operation extension, it is the responsibility of the utilities to provide the safety authorities with an exhaustive dossier demonstrating that safe extended operation is guaranteed. The role of the safety authorities is to critically evaluate the safety dossier for eventually granting the operation extension.

  1. Use of Closed Vessel as a Constant Pressure Apparatus for the Measurement of the Rate of Burning of Propellants

    Directory of Open Access Journals (Sweden)

    D. Vittal

    1980-04-01

    Full Text Available A method for the determination of burning rates of propellants whose from function is unknown is introduced. The method consists of burning in the closed vessel, a known charge weight of the test propellant alongwith a known pressure which remains nearly constant during the burning of the test propellant whose web size is the only quantity required for the evaluation of its rate of burning. The test propellants burns at near constant pressure conditions just as in the strand burner technique. This method can be applied to any unknown propellant of any shape whose web size can be measured and very large webs also can be used. In addition, the measurement of the records and the computation are very simple.

  2. Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

    Science.gov (United States)

    Krasikov, E.; Nikolaenko, V.

    2017-01-01

    Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

  3. Modeling the Ductile Brittle Fracture Transition in Reactor Pressure Vessel Steels using a Cohesive Zone Model based approach

    Energy Technology Data Exchange (ETDEWEB)

    Pritam Chakraborty; S. Bulent Biner

    2013-10-01

    Fracture properties of Reactor Pressure Vessel (RPV) steels show large variations with changes in temperature and irradiation levels. Brittle behavior is observed at lower temperatures and/or higher irradiation levels whereas ductile mode of failure is predominant at higher temperatures and/or lower irradiation levels. In addition to such temperature and radiation dependent fracture behavior, significant scatter in fracture toughness has also been observed. As a consequence of such variability in fracture behavior, accurate estimates of fracture properties of RPV steels are of utmost importance for safe and reliable operation of reactor pressure vessels. A cohesive zone based approach is being pursued in the present study where an attempt is made to obtain a unified law capturing both stable crack growth (ductile fracture) and unstable failure (cleavage fracture). The parameters of the constitutive model are dependent on both temperature and failure probability. The effect of irradiation has not been considered in the present study. The use of such a cohesive zone based approach would allow the modeling of explicit crack growth at both stable and unstable regimes of fracture. Also it would provide the possibility to incorporate more physical lower length scale models to predict DBT. Such a multi-scale approach would significantly improve the predictive capabilities of the model, which is still largely empirical.

  4. The test research on partial relieving pressure for the entry in the deep mine under high stress and friable surrounding rock

    Institute of Scientific and Technical Information of China (English)

    DU Ji-ping; HOU Chao-jiong; ZHU Ya-ping; HAO Ming-kui

    2007-01-01

    Based on the geological condition of Zhangxiaolou deep mine in Xuzhou mining area, under 986 m in depth, 20.6~31.6 MPa in maximum horizontal principal stress, and friable and fractured surrounding rock, test researches on partial relieving pressure were completed for the entry with U-steel arched yielding support. The relieving pressure parameters, technology process and results of springing blasting by boreholes and excavating pockets in the two sides of entry were introduced. It is demonstrated that springing will not be shaped under the condition of single borehole arrangement after exploded, the arrangement by a group, it will make borehole bottom form springing in 0.6~0.8 m in diameter, that convergence of two sides and roof to floor have some increments by using springing blasting for reliving pressure. This kind of method for reliving pressure is not suitable to use in the deep mine, and that the convergence of two sides obviously declined by excavating pocket in two sides, it can be still used in the entry with metal support, while maintenance of entry in deep mines is difficult, and can not be supported by bolt or bolt with wire mesh.

  5. Short range shooting distance estimation using variable pressure SEM images of the surroundings of bullet holes in textiles.

    Science.gov (United States)

    Hinrichs, Ruth; Frank, Paulo Ricardo Ost; Vasconcellos, M A Z

    2017-03-01

    Modifications of cotton and polyester textiles due to shots fired at short range were analyzed with a variable pressure scanning electron microscope (VP-SEM). Different mechanisms of fiber rupture as a function of fiber type and shooting distance were detected, namely fusing, melting, scorching, and mechanical breakage. To estimate the firing distance, the approximately exponential decay of GSR coverage as a function of radial distance from the entrance hole was determined from image analysis, instead of relying on chemical analysis with EDX, which is problematic in the VP-SEM. A set of backscattered electron images, with sufficient magnification to discriminate micrometer wide GSR particles, was acquired at different radial distances from the entrance hole. The atomic number contrast between the GSR particles and the organic fibers allowed to find a robust procedure to segment the micrographs into binary images, in which the white pixel count was attributed to GSR coverage. The decrease of the white pixel count followed an exponential decay, and it was found that the reciprocal of the decay constant, obtained from the least-square fitting of the coverage data, showed a linear dependence on the shooting distance.

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

    Science.gov (United States)

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

    2013-12-01

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

  7. Potential impact of enhanced fracture-toughness data on fracture mechanics assessment of PWR vessel integrity for pressurized thermal shock

    Energy Technology Data Exchange (ETDEWEB)

    Dickson, T.L.; Theiss, T.J.

    1991-01-01

    The Heavy Section Steel Technology (HSST) Program is involved with the generation of enhanced fracture-initiation toughness and fracture-arrest toughness data of prototypic nuclear reactor vessel steels. These two sets of data are enhanced because they have distinguishing characteristics that could potentially impact PWR pressure vessel integrity assessments for the pressurized-thermal shock (PTS) loading condition which is a major plant-life extension issue to be confronted in the 1990's. A series of large-scale fracture-mechanics experiments have produced crack-arrest (K{sub Ia}) data with the distinguishing characteristic that the values are considerably above 220 MPA {center dot} {radical}m. The implicit limit of the ASME Code and the limit used in the Integrated Pressurized Thermal Shock (IPTS) studies. Currently, the HSST Program is planning experiments to verify and quantify for A533B steel the distinguishing characteristic of elevated the distinguishing characteristic of elevated initiation-fracture toughness for shallow flaws which has been observed for other steels. The results of the analyses indicated that application of the enhanced K{sub Ia} data does reduce the conditional probability of failure P(F{vert bar}E); however, it does not appear to have the potential to significantly impact the results of PTS analyses. The application of enhanced fracture-initiation-toughness data for shallow flaws also reduces P(F{vert bar}E), and does appear to have a potential for significantly affecting the results of PTS analyses. 19 refs., 11 figs., 1 tab.

  8. Monitoring of the production quality of fibre-reinforced pressure vessels using acoustic emission testing; Ueberwachung der Fertigungsqualitaet von Faserverbund-Druckbehaeltern mittels Schallemissionspruefung

    Energy Technology Data Exchange (ETDEWEB)

    Duffner, Eric; Gregor, Christian; Bohse, Juergen [BAM Bundesanstalt fuer Materialforschung und -pruefung, Berlin (Germany)

    2011-07-01

    The investigation aimed at the validation of a test method for ensuring the production quality of reinforced-fibre pressure vessels in real fabrication conditions. The method is based on characteristics and permissible limiting values derived from acoustic emission curves during the first pressure test. The method had already been tested successfully on reinforced-fibre pressure vessels with metal liners and had been patented. With the current investigations, the possibility of detection fabrication defects in carbon fibre / glass fibre hybrid pressure vessels with polymer liners was evaluated. For this, fibre-reinforced pressure vessels were monitored by acoustic emission measurement during the first hydraulic pressure test; this test is commonly used for quality assurance of this type of pressure vessel, although without acoustic emission testing. Acoustic emission curves were registered for pressure vessels of a serial production, and the mean characteristics and their scatter were determined as reference values. These were compared with the acoustic emission curves of selectively induced fabrication defects. Fabrication defects are defects that may occur in serial production and are difficult or impossible to detect by conventional quality assurance methods. All investigated pressure vessel were then subject to stress until failure (leakage, bursting). This made it possible to verify the real influence of fabrication defects on the burst pressure and/or the fatigue characteristics of the pressure vessels and to assess the validity of acoustic emission testing. [German] Ziel der Untersuchung ist die Validierung einer Pruefmethodik zur Sicherung der Fertigungsqualitaet von Faserverbund - Druckbehaeltern unter realen Fertigungsbedingungen. Das Verfahren basiert auf Merkmalen und zulaessigen Grenzwerten, die aus Schallemissionsverlaeufen bei der Erstdruckpruefung abgeleitet werden [1]. Die Methodik konnte zuvor bereits erfolgreich an Faserverbund - Druckbehaeltern

  9. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  10. A high-throughput platform for low-volume high-temperature/pressure sealed vessel solvent extractions

    Energy Technology Data Exchange (ETDEWEB)

    Damm, Markus [Christian Doppler Laboratory for Microwave Chemistry (CDLMC) and Institute of Chemistry, Karl-Franzens-University Graz, Heinrichstrasse 28, A-8010 Graz (Austria); Kappe, C. Oliver, E-mail: oliver.kappe@uni-graz.at [Christian Doppler Laboratory for Microwave Chemistry (CDLMC) and Institute of Chemistry, Karl-Franzens-University Graz, Heinrichstrasse 28, A-8010 Graz (Austria)

    2011-11-30

    Highlights: Black-Right-Pointing-Pointer Parallel low-volume coffee extractions in sealed-vessel HPLC/GC vials. Black-Right-Pointing-Pointer Extractions are performed at high temperatures and pressures (200 Degree-Sign C/20 bar). Black-Right-Pointing-Pointer Rapid caffeine determination from the liquid phase. Black-Right-Pointing-Pointer Headspace analysis of volatiles using solid-phase microextraction (SPME). - Abstract: A high-throughput platform for performing parallel solvent extractions in sealed HPLC/GC vials inside a microwave reactor is described. The system consist of a strongly microwave-absorbing silicon carbide plate with 20 cylindrical wells of appropriate dimensions to be fitted with standard HPLC/GC autosampler vials serving as extraction vessels. Due to the possibility of heating up to four heating platforms simultaneously (80 vials), efficient parallel analytical-scale solvent extractions can be performed using volumes of 0.5-1.5 mL at a maximum temperature/pressure limit of 200 Degree-Sign C/20 bar. Since the extraction and subsequent analysis by either gas chromatography or liquid chromatography coupled with mass detection (GC-MS or LC-MS) is performed directly from the autosampler vial, errors caused by sample transfer can be minimized. The platform was evaluated for the extraction and quantification of caffeine from commercial coffee powders assessing different solvent types, extraction temperatures and times. For example, 141 {+-} 11 {mu}g caffeine (5 mg coffee powder) were extracted during a single extraction cycle using methanol as extraction solvent, whereas only 90 {+-} 11 were obtained performing the extraction in methylene chloride, applying the same reaction conditions (90 Degree-Sign C, 10 min). In multiple extraction experiments a total of {approx}150 {mu}g caffeine was extracted from 5 mg commercial coffee powder. In addition to the quantitative caffeine determination, a comparative qualitative analysis of the liquid phase coffee

  11. 联合止痛消炎膏外敷治疗血管周围皮肤肿胀疗效观察%Combination of analgesic antiphlogistic cream topical treatment curative effect observation of blood vessels surrounding skin swelling

    Institute of Scientific and Technical Information of China (English)

    张华; 张业光

    2013-01-01

    Objective To observe the combination of anti-inflammatory plaster topical treatment of fistula puncture failure within blood extravasation around blood vessels caused by swelling the clinical effect of skin. Methods 34 patients blood extravasation of xylene-induced ear edema blood vessels surrounding skin of patients, randomly divided into observation group and control group, each group of 17 cases, observation group with potato chips+xi liao completed + analgesic antiphlogistic cream topical side limb injury in fistula. In the control group with cold + 50%magnesium sulfate wet-apply internal fistula side body. Results The observation group 8 cases cured, accounted for 47%, 6 cases had marked effect, 35%, 3 cases(18%). And Control group recovered in 3 cases, accounting for 18%, 5 cases had marked effect, 29%, 7 cases (41%), and ineffective in 2 cases, accounting for 12%. Observation group of patients satisfaction 95%, control group patients' satisfaction was 80%. Observation group significantly reduce pain, swelling, eliminate time significantly shortened, subcutaneous ecchymosis dissipate quickly, no internal fistula embolism. Conclusion The combined application of analgesic antiphlogistic cream, fistula puncture failure blood extravasation to surrounding skin swelling blood vessels more effective.%目的:观察联合使用消炎止痛膏外敷治疗内瘘穿刺失败血液外渗致血管周围皮肤肿胀的临床效果。方法将34例血液外渗致血管周围皮肤肿胀的患者随机分为观察组和对照组,每组17例,观察组用土豆片+喜辽妥+止痛消炎膏外敷损伤内瘘侧肢体;对照组用冷敷+50%硫酸镁湿敷内瘘侧肢体。观察2组疗效及患者满意度情况。结果观察组痊愈8例,占47%,显效6例,占35%,有效3例,占18%;对照组痊愈3例,占18%,显效5例,占29%,有效7例,占41%,无效2例,占12%。观察组患者满意度为95%,对照组患者满意度为80%。观察组疼痛

  12. Twist seal for high-pressure vessels such as space shuttle rocket motors

    Science.gov (United States)

    von Pragenau, George L. (Inventor)

    1989-01-01

    Seals for sealing clevis and flange joints (14) of a solid rocket booster motor, and more particularly to a seal (30) which is twisted upon application of expansion forces to an edge seal (36). This twisting motion initially causes a leading edge seal (44) to be urged into sealing engagement with a surface (48) of an adjacent member (20) and thereafter, increasing fluid pressure on a pressurized side (64) of a seal (30) drives a broad sealing region (46) into sealing engagement with a surface (48).

  13. Electron density change of atmospheric-pressure plasmas in helium flow depending on the oxygen/nitrogen ratio of the surrounding atmosphere

    Science.gov (United States)

    Tomita, Kentaro; Urabe, Keiichiro; Shirai, Naoki; Sato, Yuta; Hassaballa, Safwat; Bolouki, Nima; Yoneda, Munehiro; Shimizu, Takahiro; Uchino, Kiichiro

    2016-06-01

    Laser Thomson scattering was applied to an atmospheric-pressure plasma produced in a helium (He) gas flow for measuring the spatial profiles of electron density (n e) and electron temperature (T e). Aside from the He core flow, the shielding gas flow of N2 or synthesized air (\\text{N}2:\\text{O}2 = 4:1) surrounding the He flow was introduced to evaluate the effect of ambient gas components on the plasma parameters, eliminating the effect of ambient humidity. The n e at the discharge center was 2.7 × 1021 m-3 for plasma generated with N2/O2 shielding gas, 50% higher than that generated with N2 shielding.

  14. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  15. 高压液氢容器的研制%Development of High Pressure and Liquid Hydrogen Vessel

    Institute of Scientific and Technical Information of China (English)

    路兰卿; 于洋

    2013-01-01

      针对设计温度为-253℃、设计压力为17.7 MPa高压液氢贮存容器的设计、方案确定、加工制造等方面进行了详细介绍。容器承压结构为单层厚壁板卷,其绝热方式采用液氮夹套预冷屏和外堆积绝热的形式,目前该容器已成功完成了多次发动机的试验任务。%  This text introduces a high pressure liquid hydrogen vessel’s design, plan, manufacture etc which the design temperature is-253℃ and design pressure is 17.7MPa. The pressure bearing structure is one layer, insulation is LN2 precooling screen and the outer packing insulation. This vessel is successfully designed and applied in china.

  16. 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.; Kallman, Richard A.; Maes, Miguel; Skolnik, Edward G.; Weiner, Steven C.

    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.

  17. Stress categorization in nozzle to pressure vessel connections finite elements models; Categorizacao de tensoes em modelos de elementos finitos de conexoes bocal-vaso de pressao

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Levi Barcelos de

    1999-07-01

    The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae

  18. Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.

    1979-04-01

    Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated.

  19. Influence of structural parameters on the tendency of VVER-1000 reactor pressure vessel steel to temper embrittlement

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Zabusov, O.; Fedotova, S.; Frolov, A.; Saltykov, M.; Maltsev, D.

    2013-04-01

    In this paper the influence of structural parameters on the tendency of steels to reversible temper embrittlement was studied for assessment of performance properties of reactor pressure vessel steels with extended service life. It is shown that the growth of prior austenite grain size leads to an increase of the critical embrittlement temperature in the initial state. An embrittlement heat treatment at the temperature of maximum manifestation of temper embrittlement (480 °C) shifts critical embrittlement temperature to higher values due to the increase of the phosphorus concentration on grain boundaries. There is a correlation between phosphorus concentration on boundaries of primary austenite grains and the share of brittle intergranular fracture (that, in turn, depends on impact test temperature) in the fracture surfaces of the tested Charpy specimens.

  20. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-1 and TSE-2

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.

    1976-09-01

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and two thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. The PWR calculations indicated that under some circumstances crack propagation could be expected and that experiments should be conducted for cracks that would have the potential for propagation at least halfway through the wall.

  1. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs.

  2. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi [Inner Mongolia Univ. of Science and Technology, Baotou (China). School of Material and Metallurgy; Kang, Xiaolan [Baotou Vocational and Technical College (China)

    2017-02-15

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t{sub 8/5} (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  3. Shape optimization on the nozzle of a spherical pressure vessel using the ranked bidirectional evolutionary structural optimization

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Shin; Ryu, Chung Hyun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-07-01

    To reduce stress concentration around the intersection between a spherical pressure vessel and a cylindrical nozzle under various load conditions using less material, the optimization for the distribution of reinforcement has researched. The Ranked Bidirectional Evolutionary Structural Optimization(R-BESO) method is developed recently, which adds elements based on a rank, and the performance indicator which can estimate a fully stressed model. The R-BESO method can obtain the optimum design using less iteration number than iteration number of the BESO. In this paper, the optimized intersection shape is sought using R-BESO method for a flush and a protruding nozzle. The considered load cases are a radial compression, torque and shear force.

  4. The effect of complex exercise rehabilitation program on body composition, blood pressure, blood sugar, and vessel elasticity in elderly women with obesity

    Science.gov (United States)

    Lee, Eun-Ok; Lee, Kwon-Ho; Kozyreva, Olga

    2013-01-01

    The purpose of this study is to identify what kind of effects complex exercise rehabilitation program has on body composition of female, blood pressure, blood sugar, blood vessel elasticity and find more effective complex exercise program for elderly females. The subjects are selected 30 females applicants in exercise program in City of G and not restricted in mobility to perform the exercise without any particular disorders. Exercise program is a combination of aerobic and strength training with different ratio, for the first 6 months focused on strength training complex exercise, and for next 6 months focused on aerobic exercise. Except for strength training and aerobic exercise, durations for strength, rest, and wrapping-up are equal. The frequency of experiments is 90 min each, 2 times per a week. Body composition, blood pressure, and blood vessel elasticity are tested pre and post experiment to compare the effectiveness of both complex exercises. As results, in the complex exercise program focused on strength training, weight, percent body fat, fat mass, waist hip ratio, systolic blood pressure, and diastolic pressure increased. Blood vessel elasticity maintained its level or slightly decreased. In the complex exercise focused on aerobic exercise, weight, percent body fat, fat mass, waist hip ratio, systolic pressure, and diastolic pressure decreased. Blood vessel elasticity on left foot and right foot are slightly different. Therefore, aerobic exercise is more effective than strength training for old obese females. PMID:24409428

  5. A Review of Energy Release Processes from the Failure of Pneumatic Pressure Vessels

    Science.gov (United States)

    1988-08-01

    RT) is not a good approximation. There are several equations cf state that can be used for real gases (e.g., Van der Waal’s, Beattie - Bridgeman ...The gas pressure can be written in terms of an appropriate equation of state for either an ideal or real gas. Initial fragment velocity is...assumption3 reduce Equation (1) to: -w - AE - AU (2) The ideal gas law states that foz the expailsion of a gas: W - -C, AT (3) where: C, - constant

  6. Stem Hydraulic Conductivity depends on the Pressure at Which It Is Measured and How This Dependence Can Be Used to Assess the Tempo of Bubble Pressurization in Recently Cavitated Vessels1[OPEN

    Science.gov (United States)

    Liu, Jinyu; Tyree, Melvin T.

    2015-01-01

    Cavitation of water in xylem vessels followed by embolism formation has been authenticated for more than 40 years. Embolism formation involves the gradual buildup of bubble pressure (air) to atmospheric pressure as demanded by Henry’s law of equilibrium between gaseous and liquid phases. However, the tempo of pressure increase has not been quantified. In this report, we show that the rate of pressurization of embolized vessels is controlled by both fast and slow kinetics, where both tempos are controlled by diffusion but over different spatial scales. The fast tempo involves a localized diffusion from endogenous sources: over a distance of about 0.05 mm from water-filled wood to the nearest embolized vessels; this process, in theory, should take 17 h, with complete equilibrium requiring 1 to 2 d. The implications of these timescales for the standard methods of measuring percentage loss of hydraulic conductivity are discussed in theory and deserve more research in future. PMID:26468516

  7. Studies on the Tempo of Bubble Formation in Recently Cavitated Vessels: A Model to Predict the Pressure of Air Bubbles1

    Science.gov (United States)

    Wang, Yujie; Pan, Ruihua; Tyree, Melvin T.

    2015-01-01

    A cavitation event in a vessel replaces water with a mixture of water vapor and air. A quantitative theory is presented to argue that the tempo of filling of vessels with air has two phases: a fast process that extracts air from stem tissue adjacent to the cavitated vessels (less than 10 s) and a slow phase that extracts air from the atmosphere outside the stem (more than 10 h). A model was designed to estimate how water tension (T) near recently cavitated vessels causes bubbles in embolized vessels to expand or contract as T increases or decreases, respectively. The model also predicts that the hydraulic conductivity of a stem will increase as bubbles collapse. The pressure of air bubbles trapped in vessels of a stem can be predicted from the model based on fitting curves of hydraulic conductivity versus T. The model was validated using data from six stem segments each of Acer mono and the clonal hybrid Populus 84K (Populus alba × Populus glandulosa). The model was fitted to results with root mean square error less than 3%. The model provided new insight into the study of embolism formation in stem tissue and helped quantify the bubble pressure immediately after the fast process referred to above. PMID:25907963

  8. CFD analysis of a regular sector of the ITER vacuum vessel. Part I: Flow distribution and pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Savoldi Richard, L., E-mail: laura.savoldi@polito.it [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Bonifetto, R. [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Zanino, R., E-mail: roberto.zanino@polito.it [Dipartimento Energia, Politecnico di Torino, 10129 Torino (Italy); Corpino, S.; Obiols-Rabasa, G. [Dipartimento di Ingegneria Meccanica e Aerospaziale, Politecnico di Torino, 10129 Torino (Italy); Izquierdo, J. [F4E, Barcelona (Spain); Le Barbier, R.; Utin, Y. [ITER IO, Cadarache (France)

    2013-12-15

    The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector no. 5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT{sup ®}. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10 m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ∼50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping.

  9. BIOASSAY VESSEL FAILURE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Vormelker, P

    2008-09-22

    Two high-pressure bioassay vessels failed at the Savannah River Site during a microwave heating process for biosample testing. Improper installation of the thermal shield in the first failure caused the vessel to burst during microwave heating. The second vessel failure is attributed to overpressurization during a test run. Vessel failure appeared to initiate in the mold parting line, the thinnest cross-section of the octagonal vessel. No material flaws were found in the vessel that would impair its structural performance. Content weight should be minimized to reduce operating temperature and pressure. Outer vessel life is dependent on actual temperature exposure. Since thermal aging of the vessels can be detrimental to their performance, it was recommended that the vessels be used for a limited number of cycles to be determined by additional testing.

  10. Survey of welding processes for field fabrication of 2 1/4 Cr-1 Mo steel pressure vessels. [128 references

    Energy Technology Data Exchange (ETDEWEB)

    Grotke, G.E.

    1980-04-01

    Any evaluation of fabrication methods for massive pressure vessels must consider several welding processes with potential for heavy-section applications. These include submerged-arc and shielded metal-arc, narrow-joint modifications of inert-gas metal-arc and inert-gas tungsten-arc processes, electroslag, and electron beam. The advantage and disadvantages of each are discussed. Electroslag welding can be dropped from consideration for joining of 2 1/4 Cr-1 Mo steel because welds made with this method do not provide the required mechanical properties in the welded and stress relieved condition. The extension of electron-beam welding to sections as thick as 4 or 8 inches (100 or 200 mm) is too recent a development to permit full evaluation. The manual shielded metal-arc and submerged-arc welding processes have both been employed, often together, for field fabrication of large vessels. They have the historical advantage of successful application but present other disadvantages that make them otherwise less attractive. The manual shielded metal-arc process can be used for all-position welding. It is however, a slow and expensive technique for joining heavy sections, requires large amounts of skilled labor that is in critically short supply, and introduces a high incidence of weld repairs. Automatic submerged-arc welding has been employed in many critical applications and for welding in the flat position is free of most of the criticism that can be leveled at the shielded metal-arc process. Specialized techniques have been developed for horizontal and vertical position welding but, used in this manner, the applications are limited and the cost advantage of the process is lost.

  11. HTGR Base Technology Program. Task 2: concrete properties in nuclear environment. A review of concrete material systems for application to prestressed concrete pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.

    1981-05-01

    Prestressed concrete pressure vessels (PCPVs) are designed to serve as primary pressure containment structures. The safety of these structures depends on a correct assessment of the loadings and proper design of the vessels to accept these loadings. Proper vessel design requires a knowledge of the component (material) properties. Because concrete is one of the primary constituents of PCPVs, knowledge of its behavior is required to produce optimum PCPV designs. Concrete material systems are reviewed with respect to constituents, mix design, placing, curing, and strength evaluations, and typical concrete property data are presented. Effects of extreme loadings (elevated temperature, multiaxial, irradiation) on concrete behavior are described. Finally, specialty concrete material systems (high strength, fibrous, polymer, lightweight, refractory) are reviewed. 235 references.

  12. Thermal–mechanical stress analysis of pressurized water reactor pressure vessel with/without a preexisting crack under grid load following conditions

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Soppet, William K.; Majumdar, Saurin; Natesan, Krishnamurti

    2016-12-15

    Highlights: • Use of intermittent renewable-energy source in power grid is becoming a trend. • Gird load-following can leads to variable power demand from Nuclear power plant. • Reactor components can be stressed differently under gird load-following mode. • Estimation of stress–strain state under grid load-following condition is essential. - Abstract: In this paper, we present thermal–mechanical stress analysis of a pressurized water reactor pressure vessel and its hot-leg and cold-leg nozzles. Results are presented from thermal and thermal–mechanical stress analysis under reactor heat-up, cool-down, and grid load-following conditions. Analysis results are given with and without the presence of preexisting crack in the reactor nozzle (axial crack in hot leg nozzle). From the model results it is found that the stress–strain states are significantly higher in case of presence of crack than without crack. The stress–strain state under grid load following condition are more realistic compared to the stress–strain state estimated assuming simplified transients.

  13. Heat-Induced, Pressure-Induced and Centrifugal-Force-Induced Exact Axisymmetric Thermo-Mechanical Analyses in a Thick-Walled Spherical Vessel, an Infinite Cylindrical Vessel, and a Uniform Disk Made of an Isotropic and Homogeneous Material

    Directory of Open Access Journals (Sweden)

    Vebil Yıldırım

    2017-07-01

    Full Text Available Heat-induced, pressure-induced, and centrifugal force-induced axisymmetric exact deformation and stresses in a thick-walled spherical vessel, a cylindrical vessel, and a uniform disk are all determined analytically at a specified constant surface temperature and at a constant angular velocity. The inner and outer pressures are both included in the formulation of annular structures made of an isotropic and homogeneous linear elastic material. Governing equations in the form of Euler-Cauchy differential equation with constant coefficients are solved and results are presented in compact forms. For disks, three different boundary conditions are taken into account to consider mechanical engineering applications. The present study is also peppered with numerical results in graphical forms.

  14. Main mechanisms of material properties degradation under reactor pressure vessel operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Karzov, Georgy; Timofeev, Boris [Central Research Inst. of Structural Materials ' prometey' , St. Petersburg (Russian Federation)

    1999-07-01

    In the process of NPP equipment operation materials are subjected to a prolonged influence of loads, associated with the variation of inner pressure and temperature under various conditions. Each equipment element damage is associated with some material fracture mechanism. For NPP equipment the mechanisms of irreversible damage accumulation are related with: irradiation embrittlement, thermal and strain aging, fatigue damages from mechanical and thermal loading, stress corrosion and fatigue corrosion, creep and thermal relaxation stresses, erosion and weak, thermal shock. The basic tasks of specialists working in the sphere of the provision of reliability and service life of nuclear power equipment are not only the determination of the main mechanisms of damages and reasons of their appearance, but also the study of methods which would permit to control these properties completely. By giving some examples of Russian NPP equipment with VVER-440 and VVER-1000 reactors the paper presents most typical degradation mechanisms of equipment material properties, including weldments, in the process of operation and methods to recover by using various technological means. (author)

  15. Intracranial pressure elevation reduces flow through collateral vessels and the penetrating arterioles they supply. A possible explanation for 'collateral failure' and infarct expansion after ischemic stroke.

    Science.gov (United States)

    Beard, Daniel J; McLeod, Damian D; Logan, Caitlin L; Murtha, Lucy A; Imtiaz, Mohammad S; van Helden, Dirk F; Spratt, Neil J

    2015-05-01

    Recent human imaging studies indicate that reduced blood flow through pial collateral vessels ('collateral failure') is associated with late infarct expansion despite stable arterial occlusion. The cause for 'collateral failure' is unknown. We recently showed that intracranial pressure (ICP) rises dramatically but transiently 24 hours after even minor experimental stroke. We hypothesized that ICP elevation would reduce collateral blood flow. First, we investigated the regulation of flow through collateral vessels and the penetrating arterioles arising from them during stroke reperfusion. Wistar rats were subjected to intraluminal middle cerebral artery (MCA) occlusion (MCAo). Individual pial collateral and associated penetrating arteriole blood flow was quantified using fluorescent microspheres. Baseline bidirectional flow changed to MCA-directed flow and increased by >450% immediately after MCAo. Collateral diameter changed minimally. Second, we determined the effect of ICP elevation on collateral and watershed penetrating arteriole flow. Intracranial pressure was artificially raised in stepwise increments during MCAo. The ICP increase was strongly correlated with collateral and penetrating arteriole flow reductions. Changes in collateral flow post-stroke appear to be primarily driven by the pressure drop across the collateral vessel, not vessel diameter. The ICP elevation reduces cerebral perfusion pressure and collateral flow, and is the possible explanation for 'collateral failure' in stroke-in-progression.

  16. Development of automated welding process for field fabrication of thick walled pressure vessels. Fourth quarter technical progress report for period ending September 28, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    Progress is reported in research aimed at optimizing an automated welding process for the field fabrication of thick-walled pressure vessels and for evaluating the welded joints. Information is included on the welding equipment, mechanical control of the process, joint design, filler wire optimization, in-process nondestructive testing of welds, and repair techniques. (LCL)

  17. ALICE HMPID Radiator Vessel

    CERN Multimedia

    2003-01-01

    View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

  18. Damage evaluation and analysis of composite pressure vessels using fiber Bragg gratings to determine structural health

    Science.gov (United States)

    Ortyl, Nicholas E.

    2005-11-01

    . Multiaxis fiber optic sensors are able to measure pressure, temperature, axial and transverse strain, chemical properties, corrosion, as well as transverse strain gradients. This technology is easily embedded in between the various layers of the composite structure, during manufacture, without compromising the structural integrity, in order to verify manufacturing parameters during the cure cycle and well as monitor the on-going condition of the composite structure throughout its life time. This paper reviews some of the technical work that has been accomplished during the past two years; specifically the embedding of fiber optic sensors into various composite structures in order to be able to conduct in situ non-destructive evaluation of the curing process and the service life of the component. The fiber optic technology has been developed to the point that it is at a TRL of 6.

  19. Neutron radiation embrittlement studies in support of continued operation, and validation by sampling of Magnox reactor steel pressure vessels and components

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R.B.; Bolton, C.J. [Magnox Electric plc, Berkeley Centre, Glos (United Kingdom)

    1997-02-01

    Magnox steel reactor pressure vessels differ significantly from US LWR vessels in terms of the type of steel used, as well as their operating environment (dose level, exposure temperature range, and neutron spectra). The large diameter ferritic steel vessels are constructed from C-Mn steel plates and forgings joined together with manual metal and submerged-arc welds which are stress-relieved. All Magnox vessels are now at least thirty years old and their continued operation is being vigorously pursued. Vessel surveillance and other programmes are summarized which support this objective. The current understanding of the roles of matrix irradiation damage, irradiation-enhanced copper impurity precipitation and intergranular embrittlement effects is described in so far as these influence the form of the embrittlement and hardening trend curves for each material. An update is given on the influence of high temperature exposure, and on the role of differing neutron spectra. Finally, the validation offered by the results of an initial vessel sampling exercise is summarized together with the objectives of a more extensive future sampling programme.

  20. Stress analysis in a non axisymmetric loaded reactor pressure vessel; Verificacao de tensoes em um vaso de pressao nuclear com carregamentos nao-axissimetricos

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de [Coordenadoria para Projetos Especiais (COPESP), Sao Paulo, SP (Brazil); Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

    1995-12-31

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author) 4 refs., 15 figs., 7 tabs.

  1. Evaluation of sildenafil pressurized metered dose inhalers as a vasodilator in umbilical blood vessels of chicken egg embryos.

    Science.gov (United States)

    Sawatdee, Somchai; Hiranphan, Phetai; Laphanayos, Kampanart; Srichana, Teerapol

    2014-01-01

    Sildenafil citrate is a selective phosphodiesterase-5 inhibitor used for the treatment for erectile dysfunction and pulmonary hypertension. The delivery of sildenafil directly to the lung could have several advantages over conventional treatments for pulmonary hypertension because of the local delivery, a more rapid onset of response, and reduced side effects. The major problem of sildenafil citrate is its limited solubility in water. Sildenafil citrate was complexed with cyclodextrins (CDs) to enhance its water solubility prior to development as an inhaled preparation. Four sildenafil citrate inhaled formulations were prepared with the aid of HP-β-CD (#1), α-CD (#2) and γ-CD (#3) and their effects were compared with the formulations without CDs (#4). The sildenafil citrate pressurized metered dose inhalers (pMDI) used ethanol as a solvent, PEG400 as a stabilizing agent, sorbitan monooleate as a surfactant and HFA-134a as a propellant. All formulations consisted of sildenafil citrate equivalent to a sildenafil content of 20μg/puff. These products were evaluated according to a standard guideline of inhalation products. Vasodilation testing was performed to investigate the efficacy of sildenafil pMDIs in relieving a vasoconstricted umbilical blood vessel of the chicken egg embryo. The sildenafil contents of the pMDI formulations #1-#3 were within the acceptance criteria (80-120%). The emitted doses (ED) were 102.3±11.5%, the fine particle fractions (FPF) were 60.5±5.6% and the mass median aerodynamic diameters (MMAD) were 2.3±0.3μm. The vasodilatory activity of those formulations reduced umbilical blood pressure by 67.1-73.7% after treatment by intravenous injection whereas only a 50.1-58.0% reduced blood pressure was obtained after direct spraying of the sildenafil pMDI containing CDs. With sildenafil formulations of a pMDI without CD the blood pressure was reduced by only 39.0% (P-valuevessels of chicken egg embryos after spraying sildenafil-CDs pMDIs was

  2. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  3. Investigation of a Coolant Mixing Phenomena within the Reactor Pressure Vessel of a VVER-1000 Reactor with Different Simulation Tools

    Directory of Open Access Journals (Sweden)

    V. Sánchez

    2010-01-01

    Full Text Available The Institute of Neutron Physics and Reactor Technology (INR is involved in the qualification of coupled codes for reactor safety evaluations, aiming to improve their prediction capability and acceptability. In the frame of the VVER-1000 Coolant Transient Benchmark Phase 1, RELAP5/PARCS has been extensively assessed. Phase 2 of this benchmark was focused on both multidimensional thermal hydraulic phenomena and core physics. Plant data will be used to qualify the 3D models of TRACE and RELAP5/CFX, which were coupled for this purpose. The developed multidimensional models of the VVER-1000 reactor pressure vessel (RPV as well as the performed calculations will be described in detail. The predicted results are in good agreement with experimental data. It was demonstrated that the chosen 3D nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. Even though only a 3D coarse nodalization is used in TRACE, the integral results are comparable to those obtained by RELAP5/CFX.

  4. Characterisation of interfacial segregation to Cu-enriched precipitates in two thermally aged reactor pressure vessel steel welds

    Energy Technology Data Exchange (ETDEWEB)

    Styman, P.D., E-mail: paul.styman@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); National Nuclear Laboratory, B168, Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Hyde, J.M. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); National Nuclear Laboratory, B168, Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Wilford, K.; Parfitt, D.; Riddle, N. [Rolls-Royce, PO Box 2000, Raynesway, Derby DE21 7XX (United Kingdom); Smith, G.D.W. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom)

    2015-12-15

    To understand the contribution of long term thermal ageing to Reactor Pressure Vessel (RPV) embrittlement two high Cu steel welds with different Ni contents were thermally aged for times up to 100,000 h at 330 °C and 365 °C. Microstructural characterisation using Atom Probe Tomography was performed. Thermal ageing produced a high number density of nano-scale Cu-enriched precipitates. The precipitate–matrix interfaces were enriched in Ni, Mn and Si. The characterisation of these interfaces using a double cluster search approach is the subject of this work. The interface region around thermally-induced precipitates was found to be wider in steels with higher bulk Ni contents and where precipitates had larger core radii. The effect of ageing temperature on interface width was small when comparing precipitates of equal core radius. The narrower interface width in the lower Ni steels is reflected in the composition of the interface, which has a lower Ni content than in the higher Ni material. The reduction in interfacial energy due to the segregation of Ni, Mn and Si has been calculated and shows enhanced reductions in interfacial energy with increasing precipitate size, but no obvious effect of temperature. - Highlights: • Characterisation of interfacial segregation of Ni, Mn and Si to Cu-enriched clusters. • Analysis method gives information on interface composition and widths of large numbers of clusters. • Reduction in interface energy due to segregation of Ni, Mn and Si is calculated.

  5. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  6. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su, E-mail: camila@lasme.coppe.ufrj.b, E-mail: gabrielromero@lasme.coppe.ufrj.b, E-mail: sujian@lasme.coppe.ufrj.b [Universidade Federal do Rio de Janeiro (COPPE/UFRJ), RJ (Brazil). Nuclear Engineering Program

    2010-07-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-{omega} based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  7. The cryogenic bonding evaluation at the metallic-composite interface of a composite overwrapped pressure vessel with additional impact investigation

    Science.gov (United States)

    Clark, Eric A.

    A bonding evaluation that investigated the cryogenic tensile strength of several different adhesives/resins was performed. The test materials consisted of 606 aluminum test pieces adhered to a wet-wound graphite laminate in order to simulate the bond created at the liner-composite interface of an aluminum-lined composite overwrapped pressure vessel. It was found that for cryogenic applications, a flexible, low modulus resin system must be used. Additionally, the samples prepared with a thin layer of cured resin -- or prebond -- performed significantly better than those without. It was found that it is critical that the prebond surface must have sufficient surface roughness prior to the bonding application. Also, the aluminum test pieces that were prepared using a surface etchant slightly outperformed those that were prepared with a grit blast surface finish and performed significantly better than those that had been scored using sand paper to achieve the desired surface finish. An additional impact investigation studied the post impact tensile strength of composite rings in a cryogenic environment. The composite rings were filament wound with several combinations of graphite and aramid fibers and were prepared with different resin systems. The rings were subjected to varying levels of Charpy impact damage and then pulled to failure in tension. It was found that the addition of elastic aramid fibers with the carbon fibers mitigates the overall impact damage and drastically improves the post-impact strength of the structure in a cryogenic environment.

  8. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  9. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.

    2005-06-01

    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  10. Subaquatic, pressure vessels and LPG storage spheres internal inspection; Inspecao interna de esfera utilizando mergulho como acesso

    Energy Technology Data Exchange (ETDEWEB)

    Filgueira Filho, Rafael; Monteiro, Ayres [PETROBRAS, Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Minimizing shut-down costs is a widespread target in the oil and gas industry. The use of new inspection techniques is one of the ways for that. This work presents a new procedure for internal inspections in pressure vessels by the non destructive testing - NDT, ACFM, using industrial diving techniques. As a pioneer experience, this method was applied in the inspection of the internal parts of the LPG sphere tank 5101 at PETROBRAS Transporte S.A. - TRANSPETRO, in Jequie's Terminal, in the state of Bahia, in december, 2003. This new method allows the reduction of indirect costs related to operational unavailability of the equipment, by the reduction of the shut-down time in approximately 50%, when compared to the demanded shut down time, when using scaffolds for accessing the internal parts. Despite of direct costs are still higher with the new methodology, this paper demonstrates the economical feasibility of this new method, based on the savings obtained with the fastest return of the equipment to operation. (author)

  11. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Tricot, N. [Institut de Radioprotection et de Surete Nucleaire, IRSN/DES/SECCA, 92 - Fontenay aux Roses (France); Jendrich, U. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  12. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kameda, J. [National Institute for Materials Science, Sengen, Tsukuba 305-0047 (Japan); Nagai, Y.; Toyama, T.; Matsukawa, Y. [Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Onizawa, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2012-06-15

    The effect of thermal aging on microstructural changes in stainless steel submerged arc weld-overlay cladding of reactor pressure vessels was investigated using atom probe tomography (APT). In as-received materials subjected to post-welding heat treatments (PWHTs), with a subsequent furnace cooling, a slight fluctuation of the Cr concentration was observed due to spinodal decomposition in the {delta}-ferrite phase but not in the austenitic phase. Thermal aging at 400 Degree-Sign C for 10,000 h caused not only an increase in the amplitude of spinodal decomposition but also the precipitation of G phases with composition ratios of Ni:Si:Mn = 16:7:6 in the {delta}-ferrite phase. The degree of the spinodal decomposition in the submerged arc weld sample was similar to that in the electroslag weld one reported previously. We also observed a carbide on the {gamma}-austenite and {delta}-ferrite interface. There were no Cr depleted zones around the carbide.

  13. Development and evaluation of thermoelectric power measurements as a non destructive technique to evaluate ageing of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Acosta, B.; Debarberis, L. [European Commission, JRC Institute for Advanced Materials, Petten (Netherlands); Perlado, J.M. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM, E.T.S.I.I., Madris (Spain)

    2001-07-01

    The STEAM (Seebeck and Thomson effects on aged materials) technique developed at the JRC-IAM (joint research centre - institute for advanced materials), is a new non-destructive method able to detect in a simple way degradation of materials, in particular to be applied on those steels that form the reactor pressure vessel of nuclear plants. The STEAM method is based on the measurement of the thermoelectric voltage generated by the Seebeck and Thomson effects taking place in the material under test. In order to evaluate the performance of the STEAM technique on irradiated material a set of 32 model alloys was selected. Measurements with the STEAM technique have been performed on the model alloys in both conditions, fresh and irradiated, with the aim of correlating the irradiation induced embrittlement and the change on the Seebeck coefficient due to irradiation. The results show that there is a relationship between transition temperature shifts and Seebeck coefficient value change between irradiated and fresh materials. In order to understand the response of the Seebeck coefficient to neutron irradiation damage a model based on multivariable correlation analysis is proposed. (A.C.)

  14. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)

    2016-11-15

    This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  15. Effects of Geometry of Reactor Pressure Vessel Upper Head Control Rod Drive Mechanism Penetration Nozzles on J-Groove Weld Residual Stress

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Hee; Kim, Yun Jae; Bae, Hong Yeol [Korea University, Seoul (Korea, Republic of); and others

    2011-10-15

    In pressurized water reactors (PWRs), the reactor pressure vessel (RPV) upper head contains numerous control rod drive mechanism (CRDM) nozzles. In the last 10 years, the incidences of cracking in alloy 600 CRDM nozzles and their associated welds has increased significantly. Several axial and circumferential cracks have been found in CRDM nozzles in European PWRs and U.S. nuclear power plants. These cracks are caused by primary water stress corrosion cracking (PWSCC) and have been shown to be driven by welding residual stresses and operational stresses in the weld region. Therefore, detailed finite-element (FE) simulations for the Korea Nuclear Reactor Pressure Vessel have been conducted in order to predict the magnitudes of the weld residual stresses in the tube materials. In particular, the weld residual stress results are compared in terms for nozzle location, geometry factor ro/t, geometry of fillet, and adjacent nozzle.

  16. News from the Library: A new key reference work for the engineer: ASME's Boiler and Pressure Vessel Code at the CERN Library

    CERN Multimedia

    CERN Library

    2011-01-01

    The Library is aiming at offering a range of constantly updated reference books, to cover all areas of CERN activity. A recent addition to our collections strengthens our offer in the Engineering field.   The CERN Library now holds a copy of the complete ASME Boiler and Pressure Vessel Code, 2010 edition. This code establishes rules of safety governing the design, fabrication, and inspection of boilers and pressure vessels, and nuclear power plant components during construction. This document is considered worldwide as a reference for mechanical design and is therefore important for the CERN community. The Code published by ASME (American Society of Mechanical Engineers) is kept current by the Boiler and Pressure Committee, a volunteer group of more than 950 engineers worldwide. The Committee meets regularly to consider requests for interpretations, revision, and to develop new rules. The CERN Library receives updates and includes them in the volumes until the next edition, which is expected to ...

  17. Site Measurement of Surrounding Rock Pressure and Analysis of Structure Stress of Large-span Bias-pressured Shallow Tunnels%大跨浅埋偏压隧道围压实测及结构受力分析

    Institute of Scientific and Technical Information of China (English)

    江磊; 侯哲生; 吴海卫

    2015-01-01

    大跨浅埋偏压隧道由于其非对称的受力条件,易引发结构变形与开裂等突出问题,近年来越来越受工程技术界的重视。以邢汾高速公路邢台段后偏梁大跨度隧道浅埋偏压段为例,通过实测围岩压力,根据荷载结构法的基本原理,利用 ANSYS 软件对其结构受力特性进行数值分析,得到结论:实测的围岩压力分布和偏压的地形之间具有较为一致的对应性,即围岩压力受地形的影响显著;二衬总应力受轴力引起的应力影响较小,主要受弯矩引起的应力控制,整个拱圈范围内最危险的部位是在受围岩压力最大的左侧拱肩处;随着二衬厚度的变化,二衬总应力在不同的部位均发生相应变化,但变化幅度均不大。相关研究结论为后续类似大跨浅埋偏压隧道的合理设计与施工提供参考依据。%The asymmetrical stress condition of the Large-span bias-pressured shallow tunnels can cause many problems easily ,such as structural deformation and cracking .These problems have attracted more and more attentions from engi-neering and technology research field in recent years .The Houpianliang large-span bias-pressured shallow tunnel in Xing-tai section of Xingtai - Fenyang highway was taken as an example to numerically analyze the mechanical characteristics of the structure by adopting ANSYS ,according to the data of site measurement of surrounding rock pressure and the princi -ple of load structure method .The results indicate that :the actual measured surrounding rock pressure is consistent with the bias terrain ,which means the surrounding rock pressure is influenced by topography significantly ;the total stress of the second lining is mainly controlled by the stress from bending moment and is less influenced by the stress from axial force ,the weakest part within the scope of the arch ring is the left spandrel which is subject to maximum pressure ;with the change of the

  18. The application of ultrasonic testing technology of pressure vessels.%压力容器超声检测技术的应用

    Institute of Scientific and Technical Information of China (English)

    2013-01-01

    Pressure vessels are widely used in chemical,polymer physics and other fields,because all needs to run in under extreme conditions,such as high pressure high temperature,the pressure vessel manufacture,use process requirements are higher,nondestructive testing technology is to guarantee the safe operation of the production and one of the important methods of product quality and reliable. Pressure vessel in the manufacturing process,rely mainly on ultrasonic testing this paper mainly introduces characteristics and applicable scope of ultrasonic testing technology,in order to better work convenient.%  压力容器被广泛地应用于化工、高分子物理等领域,因均需在高压高温等极端条件下运行,对压力容器的制造、使用过程要求都较高,无损检测技术是保证生产安全运行和产品质量可靠的重要方法之一。压力容器在制造过程中,主要依靠超声进行检测本文主要介绍超声检测技术的特点及适用范围,以更好的方便开展工作。

  19. Experimental study on the pressure and pulse wave propagation in viscoelastic vessel tubes-effects of liquid viscosity and tube stiffness.

    Science.gov (United States)

    Ikenaga, Yuki; Nishi, Shohei; Komagata, Yuka; Saito, Masashi; Lagrée, Pierre-Yves; Asada, Takaaki; Matsukawa, Mami

    2013-11-01

    A pulse wave is the displacement wave which arises because of ejection of blood from the heart and reflection at vascular bed and distal point. The investigation of pressure waves leads to understanding the propagation characteristics of a pulse wave. To investigate the pulse wave behavior, an experimental study was performed using an artificial polymer tube and viscous liquid. A polyurethane tube and glycerin solution were used to simulate a blood vessel and blood, respectively. In the case of the 40 wt% glycerin solution, which corresponds to the viscosity of ordinary blood, the attenuation coefficient of a pressure wave in the tube decreased from 4.3 to 1.6 dB/m because of the tube stiffness (Young's modulus: 60 to 200 kPa). When the viscosity of liquid increased from approximately 4 to 10 mPa·s (the range of human blood viscosity) in the stiff tube, the attenuation coefficient of the pressure wave changed from 1.6 to 3.2 dB/m. The hardening of the blood vessel caused by aging and the increase of blood viscosity caused by illness possibly have opposite effects on the intravascular pressure wave. The effect of the viscosity of a liquid on the amplitude of a pressure wave was then considered using a phantom simulating human blood vessels. As a result, in the typical range of blood viscosity, the amplitude ratio of the waves obtained by the experiments with water and glycerin solution became 1:0.83. In comparison with clinical data, this value is much smaller than that seen from blood vessel hardening. Thus, it can be concluded that the blood viscosity seldom affects the attenuation of a pulse wave.

  20. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  1. Short-Term Blood Pressure Variability Relates to the Presence of Subclinical Brain Small Vessel Disease in Primary Hypertension.

    Science.gov (United States)

    Filomena, Josefina; Riba-Llena, Iolanda; Vinyoles, Ernest; Tovar, José L; Mundet, Xavier; Castañé, Xavier; Vilar, Andrea; López-Rueda, Antonio; Jiménez-Baladó, Joan; Cartanyà, Anna; Montaner, Joan; Delgado, Pilar

    2015-09-01

    Blood pressure (BP) variability is associated with stroke risk, but less is known about subclinical cerebral small vessel disease (CSVD). We aimed to determine whether CSVD relates to short-term BP variability independently of BP levels and also, whether they improve CSVD discrimination beyond clinical variables and office BP levels. This was a cohort study on asymptomatic hypertensives who underwent brain magnetic resonance imaging and 24-hour ambulatory BP monitoring. Office and average 24-hour, daytime and nighttime BP levels, and several metrics of BP variability (SD, weighted SD, coefficient of variation, and average real variability [ARV]) were calculated. Definition of CSVD was based on the presence of lacunar infarcts and white matter hyperintensity grades. Multivariate analysis and integrated discrimination improvement were performed to assess whether BP variability and levels were independently associated with CSVD and improved its discrimination. Four hundred eighty-seven individuals participated (median age, 64; 47% women). CSVD was identified in 18.9%, related to age, male sex, diabetes mellitus, use of treatment, ambulatory BP monitoring-defined BP levels, and ARV of systolic BP at any period. The highest prevalence (33.7%) was found in subjects with both 24-hour BP levels and ARV elevated. BP levels at any period and ARV (24 hours and nocturnal) emerged as independent predictors of CSVD, and discrimination was incrementally improved although not to a clinically significant extent (integrated discrimination improvement, 5.31%, 5.17% to 5.4%). Ambulatory BP monitoring-defined BP levels and ARV of systolic BP relate to subclinical CSVD in hypertensive individuals.

  2. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf (Germany)

    2008-07-01

    WWER-440 second generation (V-213) reactor pressure vessels (RPV) were produced by IZHORA in Russia and by SKODA in the former Czechoslovakia. The surveillance Charpy-V and fracture mechanics SE(B) specimens of both producers have different orientations. The main difference is the crack extension direction which is through the RPV thickness and circumferential for ISHORA and SKODA RPV, respectively. In particular for the investigation of weld metal from multilayer submerged welding seams the crack extension direction is of importance. Depending on the crack extension direction in the specimen there are different welding beads or a uniform structure along the crack front. The specimen orientation becomes more important when the fracture toughness of the weld metal is directly determined on surveillance specimens according to the Master Curve (MC) approach as standardised in the ASTM Standard Test Method E1921. This approach was applied on weld metal of the RPV beltline welding seam of Greifswald Unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The specimens are in TL and TS orientation. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the MC. Nearly all values lie within the fracture toughness curves for 5% and 95% fracture probability. There is a strong variation of the reference temperature T{sub 0} though the thickness of the welding seam, which can be explained with structural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TS and TL orientation in the welding seam have a differentiating and integrating behaviour, respectively. The statistical assumptions behind the MC approach are valid for both specimen orientations even if the structure is not uniform along the crack front. By comparison crack extension, JR, curves measured on SE(B) specimens with TL and TS orientation

  3. Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Lavrukhina, Z. V.; Saltykov, M. A.; Fedotova, S. V.; Khodan, A. N.

    2016-08-01

    In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20-71)·1022 m-2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

  4. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    Science.gov (United States)

    Oh, Yong-Jun; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T0 values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 μm in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness ( KJC) was inversely proportional to the square root of the triggering inclusion diameter ( di) at respective temperatures. From this relationship, we determined median KJC values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median KJC ( K'JC(med) ). The obtained K'JC(med) values showed quite smaller deviation from the master curve at different temperatures than the experimental median KJC values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T0 value.

  5. Microstructural parameters governing cleavage fracture behaviors in the ductile-brittle transition region in reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won-Jon; Lee, Bong-Sang; Oh, Yong-Jun; Huh, Moo-Young; Hong, Jun-Hwa

    2004-08-15

    The fracture behaviors in the ductile-brittle transition region of reactor pressure vessel (RPV) steels with similar chemical compositions but different manufacturing processes were examined in view of cleavage fracture stress at crack-tip. The steels typically had a variation in grain size and carbide size distribution through the different manufacturing processes. Fracture toughness was evaluated by using a statistical method in accordance to the ASTM standard E1921. From the fractography of the tested specimens, it was found that fracture toughness of the steels increased with increasing distance from the crack-tip to the cleavage initiating location, namely cleavage initiation distance (CID, X{sub f}) and its statistical mean value (K{sub JC(med)}) was proportional to the cleavage fracture stress ({sigma}{sub f}) determined from finite-element (FE) calculation at cleavage initiating location. On the other hand, {sigma}{sub f} could also be calculated by applying the size of microstructural parameters, such as carbide, grain and bainite packet, into the Griffith's theory for brittle fracture. Among the parameters, the {sigma}{sub f} obtained from the mean diameter of the carbides above 1% of the total population was in good agreement with the {sigma}{sub f} value from the FE calculation for the five different steels. The results suggest that the fracture toughness of bainitic RPV steels in the transition region is mostly influenced by only some 1% of total carbides and the critical step for cleavage fracture of the RPV steels should be the propagation of this carbide size crack to the adjacent ferrite matrix.

  6. The effect of non-metallic inclusions on the fracture toughness master curve in high copper reactor pressure vessel welds

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Yong-Jun E-mail: yjoh@kaeri.re.kr; Lee, Bong-Sang; Hong, Jun-Hwa

    2002-03-01

    The fracture toughness of two high copper reactor pressure vessel welds having low upper shelf energy was evaluated in accordance with the master curve method of ASTM E1921. The resultant data were correlated to the metallurgical factors involved in the brittle fracture initiation to provide a metallurgical-based understanding of the master curve. The tests were performed using pre-cracked Charpy V-notched specimens and the master curve was made with an average of T{sub 0} values determined at different temperatures. In all specimens, the cleavage fracture initiated at non-metallic inclusion ranging from 0.7 to 3.5 {mu}m in diameter showing a scatter with the specimens and testing temperatures. Temperature dependency of the triggering particle size was not found. The fracture toughness (K{sub J{sub C}}) was inversely proportional to the square root of the triggering inclusion diameter (d{sub i}) at respective temperatures. From this relationship, we determined median K{sub J{sub C}} values which correspond to the average value of triggering inclusion diameter of all tested specimens and defined them as a modified median K{sub J{sub C}} (K{sup '}{sub J{sub C}}{sub (med)}). The obtained K{sup '}{sub J{sub C}}{sub (med)} values showed quite smaller deviation from the master curve at different temperatures than the experimental median K{sub J{sub C}} values. This suggests that the master curve is on the premise of a constant dimension of key microstructural factor in a material regardless of the testing temperature. But the inclusion size at trigger point played an important role in the absolute position of the master curve with temperature and the consequent T{sub 0} value.

  7. Cyclic Crack Growth Testing of an A.O. Smith Multilayer Pressure Vessel with Modal Acoustic Emission Monitoring and Data Assessment

    Science.gov (United States)

    Ziola, Steven M.

    2014-01-01

    Digital Wave Corp. (DWC) was retained by Jacobs ATOM at NASA Ames Research Center to perform cyclic pressure crack growth sensitivity testing on a multilayer pressure vessel instrumented with DWC's Modal Acoustic Emission (MAE) system, with captured wave analysis to be performed using DWCs WaveExplorerTM software, which has been used at Ames since 2001. The objectives were to document the ability to detect and characterize a known growing crack in such a vessel using only MAE, to establish the sensitivity of the equipment vs. crack size and / or relevance in a realistic field environment, and to obtain fracture toughness materials properties in follow up testing to enable accurate crack growth analysis. This report contains the results of the testing.

  8. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident; Reaktoripaineastian pohjan ja laepivientien kuumeneminen sydaemen sulamisonnettomuudessa

    Energy Technology Data Exchange (ETDEWEB)

    Ikonen, K. [VTT Energy, Espoo (Finland). Nuclear Energy

    1997-10-01

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.).

  9. 核电压力容器焊接过程中的生产管理%Production Control to Welding of Nuclear Reactor Pressure Vessels

    Institute of Scientific and Technical Information of China (English)

    宋桂艳

    2014-01-01

    The article describes the welding method for nuclear reactor pressure vessels and the key points of plan control and welding production control and process control in welding operation.%概述核电压力容器焊接的特点,指出焊接生产过程中计划管理、焊接生产过程管理和过程控制的重点内容。

  10. 高温、高压、临氢在用压力容器 的氢腐蚀检验%HYDROGEN ATTACK INSPECTION ON HIGH PRESSURE VESSEL IN SERVICE UNDER HIGH TEMPERATURE,HIGH PRESSURE AND HYDROGEN ENVIROMENT

    Institute of Scientific and Technical Information of China (English)

    乔学福

    2001-01-01

    The material property of pressure vessel in high temperature,high pressure and hydrogen service can be clanged chaused by hydrogen attack.In combination with the main material inspection of 123-c heat exchanger in synthetic ammonia unit put into production for more than 20 years,the key paints and mehtod of inspecton on pressure vessel in service under high temperature,high pressure and hydrogen serice are mentioned,the mechanisum of hydrogen attack is analyzed.The determination of safety grade for hydrogen attacked pressure vessel in serice is given by the author with personal viewpoint.%高温、高压、临氢压力容器,由于氢的腐蚀其材质的性能会发生变化。作者结合对投用20多年的合成氨装置123—C换热器主体材质的检验,阐述了高温、高压、临氢在用压力容器的检验要点和方法,分析了氢腐蚀产生的机理,对已发生了氢腐蚀的在用压力容器的安全等级判定提出了个人的看法。

  11. Vessel Operating Units (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

  12. [Development and Validation of a Fully Automated, Experimental Set-Up for Ex-Vivo Burst Pressure Testing after Surgical Vessel Closure].

    Science.gov (United States)

    Wallimann, Herbert; Menges, Pia; Hausen, Bernard; Linder, Albert

    2017-06-20

    Background A growing number of operations are performed using minimally invasive techniques. Therefore, a lot of new requirements must be met by the staplers currently available. At the present time, the most widely used methods of minimally invasive vascular occlusion involve high-frequency energy, clips, and staplers. The most important quality parameter is burst pressure, which is measured with a variety of experimental set-ups, all of which are subject to criticism. With this study, we want to introduce a fully automated vascular burst pressure measuring system that largely mimics physiological conditions. An important feature of this set-up is the detection of very early leakage from the staple line (FAIR Leakage = First Appearance of Leakage requiring Intervention). Material and Methods Burst pressure was measured in vessel segments of porcine common carotid arteries. For vascular occlusion, we used the stapler device Micro Cutter XCHANGE(®) by DexteraSurgical. Prior to closure, the vessel was filled to a pressure of 80 mmHg. The pressure was increased at a defined flow rate. Burst pressure was defined as staple line leakage requiring intervention. Results and Validation 30 staple lines were examined. The average burst pressure visually determined by two independent investigators was 515.8 mmHg ± 236.3 mmHg. Maximal burst pressure was 911 mmHg, and minimal burst pressure 80 mmHg. The average burst pressure detected electronically was 511.8 mmHg ± 239.1 mmHg. Statistically, there was a highly significant correlation of visually and electronically detected burst pressures. Conclusion This is the first experimental set-up for a systematic burst pressure test that is fully automated and therefore eliminates any bias related to the investigator. The experimental set-up with a defined intravascular pressure prior to closure and the use of a liquid with blood-like viscosity enabled us to largely mimic intraoperative conditions. Since burst

  13. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  14. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  15. Development of Mini-Compact Tension Test Method for Determining Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Validation of the mini compact tension specimen (mini-CT) geometry has been performed on previously well characterized Midland beltline Linde 80 (WF-70) weld in the unirradiated condition. It was shown that the fracture toughness transition temperature, To, measured by these Mini-CT specimens is almost the same as To value that was derived from various larger fracture toughness specimens. Moreover, an International collaborative program has been established to extend the assessment and validation efforts to irradiated Linde 80 weld metal. The program is underway and involves the Oak Ridge National Laboratory (ORNL), Central Research Institute for Electrical Power Industry (CRIEPI), and Electric Power Research Institute (EPRI). The irradiated Mini-CT specimens from broken halves of previously tested Charpy

  16. Fracture mechanics characterisation of the WWER-440 reactor pressure vessel beltline welding seam of Greifswald unit 8

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner; Schuhknecht, Jan [Forschungszentrum Dresden-Rossendorf, Dresden (Germany)

    2009-07-01

    Russian type WWER reactors are operated in Russia and many other European countries like Finland, Czech Republic, Slovak Republic, Hungary, Bulgaria and Ukraine. Surveillance specimen programmes for the inspection of the aging of the reactor pressure vessel (RPV) materials were implemented for the second generation of WWER-440/V-213 reactors. The test results and the RPV integrity assessment has been evaluated according to national codes based on the Russian code PNAE G-7-002-86 ''Strength Calculation Norms for Nuclear Power Plant Equipment and Piping'' [1]. This is an indirect and correlative approach of determining the fracture toughness of the RPV steels in the initial and irradiated condition. The Master Curve (MC) approach as adopted in the test procedure ASTM E1921 [2] for assessing the fracture toughness of sampled irradiated materials has been gaining acceptance throughout the world [3]. The MC approach is more naturally suited to probabilistic analyses because it defines both a mean transition toughness value and a distribution around that value. It contains the assumptions of macroscopically homogenous material with uniform distribution of crack initiating defects along the crack front. In contrast to present indirect and correlative approach the specimen orientation and especially the crack extension direction in multilayer weld metal becomes more important for the direct measurement of the fracture toughness with Charpy size SE(B) specimens. The orientation of the Charpy- and SE(B) specimens is different for RPVs manufactured in Russia and by the SKODA company in the former Czechoslovakia [4,5]. Particularly with regard to weld metal it can be expected that the parameters of fracture toughness measured with Charpy-V or SE(B) specimens are strongly influenced by the specimen orientation. It raises the question whether the MC approach is also applicable when the structure varies along the crack front which is happen in TL oriented SE

  17. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  18. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  19. Program to develop acoustic emission-flaw relationship for inservice monitoring of nuclear pressure vessels. Annual report, July 1, 1976 - October 1, 1977

    Energy Technology Data Exchange (ETDEWEB)

    Hutton, P.H.; Kurtz, R.J.; Schwenk, E.B.; Pavloff, C.

    1978-06-01

    Laboratory mechanical tests were conducted to evaluate AE during uniaxial tensile, fracture and fatigue crack growth in A533B pressure vessel steel. The A533B steel included two heats of Class 1, one heat of Class 2 and a weldment made for the Heavy Section Steel Technology (HSST) Program. Specimen types included uniaxial tensile specimens, size 2 compact tension specimens for fatigue crack growth and fracture tests, and a single-edge notch specimen also for fatigue crack growth through material that was uniformly strained 3% prior to fatigue testing. In addition, AE monitoring was conducted on the HSST V-7B 6-inch thick pressure vessel test. AE data were partitioned into four ranges of signal amplitude and rise time. All the AE data were analyzed, with respect to mechanical behavior of A533B steel. Linear elastic fracture mechanics analysis methods were used to relate AE parameters to fracture and fatigue crack growth parameters. AE data from the V-7B vessel test were correlated with stress intensity factor and crack opening displacement. AE data from the fatigue crack growth tests were investigated using models based on fatigue crack growth rate, fatigue crack area and theoretical crack tip plastic zone size.

  20. Confinement Vessel Dynamic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    R. Robert Stevens; Stephen P. Rojas

    1999-08-01

    A series of hydrodynamic and structural analyses of a spherical confinement vessel has been performed. The analyses used a hydrodynamic code to estimate the dynamic blast pressures at the vessel's internal surfaces caused by the detonation of a mass of high explosive, then used those blast pressures as applied loads in an explicit finite element model to simulate the vessel's structural response. Numerous load cases were considered. Particular attention was paid to the bolted port connections and the O-ring pressure seals. The analysis methods and results are discussed, and comparisons to experimental results are made.

  1. Analysis of dosimetry from the H.B. Robinson unit 2 pressure vessel benchmark using RAPTOR-M3G and ALPAN

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.A. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2011-07-01

    Document available in abstract form only, full text of document follows: The dosimetry from the H. B. Robinson Unit 2 Pressure Vessel Benchmark is analyzed with a suite of Westinghouse-developed codes and data libraries. The radiation transport from the reactor core to the surveillance capsule and ex-vessel locations is performed by RAPTOR-M3G, a parallel deterministic radiation transport code that calculates high-resolution neutron flux information in three dimensions. The cross-section library used in this analysis is the ALPAN library, an Evaluated Nuclear Data File (ENDF)/B-VII.0-based library designed for reactor dosimetry and fluence analysis applications. Dosimetry is evaluated with the industry-standard SNLRML reactor dosimetry cross-section data library. (authors)

  2. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    Science.gov (United States)

    van Duysen, J. C.; Meric de Bellefon, G.

    2017-02-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  3. Pressure and temperature profile data collected by the NOAA vessel Bay Hydrographer during survey operations along the NE US coast, 03 February 2005 to 21 November 2005 (NODC Accession 0002670)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Pressure and temperature profile data were collected using CTD casts from the NOAA Survey Vessel BAY HYDROGRAPHER. Data were collected in the Chesapeake Bay from...

  4. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  5. A new system for cutting brain tissue preserving vessels: water jet cutting.

    Science.gov (United States)

    Terzis, A J; Nowak, G; Rentzsch, O; Arnold, H; Diebold, J; Baretton, G

    1989-01-01

    The water jet cutting system allows transaction and dissection of biological structures with little bleeding. Structures of higher tissue rigidity remain unchanged while softer tissues are mechanically dissected. In brain tissue, all vessels larger than 20 microns are left intact after the passage of the jet stream with a pressure of up to 5 bar, and therefore vessels can be isolated selectively from the surrounding tissue. Oedema is present adjacent to the cut and no increase of temperature occurs.

  6. Dust explosions in spherical vessels: prediction of the pressure evolution and determination of the burning velocity and flame thickness

    NARCIS (Netherlands)

    Dahoe, A.E.; Zevenbergen, J.F.; Verheijen, P.J.T.; Lemkowitz, S.M.; Scarlett, B.

    1996-01-01

    A well known limitation of the ’cube-root-law’ is that it becomes invalid when the flame thickness is significant with respect to the vessel radius. In the literature flame thicknesses in dust-air mixtures ranging from 15 to 80 centimeters have been reported [1], which exceed the radii of the

  7. Dust explosions in spherical vessels: prediction of the pressure evolution and determination of the burning velocity and flame thickness

    NARCIS (Netherlands)

    Dahoe, A.E.; Zevenbergen, J.F.; Verheijen, P.J.T.; Lemkowitz, S.M.; Scarlett, B.

    1996-01-01

    A well known limitation of the ’cube-root-law’ is that it becomes invalid when the flame thickness is significant with respect to the vessel radius. In the literature flame thicknesses in dust-air mixtures ranging from 15 to 80 centimeters have been reported [1], which exceed the radii of the 20-1it

  8. Effects of Repair Weld of Reactor Pressure Vessel Upper Head Control Rod Drive Mechanism Penetration Nozzle on J-Groove Weldment Using Finite Element Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Hee; Yoo, Sam Hyeon [Korea Military Academy, Seoul (Korea, Republic of); Kim, Yun Jae [Korea University, Seoul (Korea, Republic of)

    2014-06-15

    In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.

  9. Effect of Initial Moisture Content on the in-Vessel Composting Under Air Pressure of Organic Fraction of MunicipalSolid Waste in Morocco

    Directory of Open Access Journals (Sweden)

    Abdelhadi Makan

    2013-01-01

    Full Text Available This study aimed to evaluate the effect of initial moisture content on the in-vessel composting under air pressure of organic fraction of municipal solid waste in Morocco in terms of internal temperature, produced gases quantity, organic matter conversion rate, and the quality of the final composts.For this purpose, in-vessel bioreactor was designed and used to evaluate both appropriate initial air pressure and appropriate initial moisture content for the composting process. Moreover, 5 experiments were carried out within initial moisture content of 55%, 65%, 70%, 75% and 85%. The initial air pressure and the initial moisture content of the mixture showed a significant effect on the aerobic composting. The experimental results demonstrated that for composting organic waste, relatively high moisture contents are better at achieving higher temperatures and retaining them for longer times.This study suggested that an initial moisture content of around 75%, under 0.6 bar, can be considered as being suitable for efficient composting of organic fraction of municipal solid waste. These last conditions, allowed maximum value of temperature and final composting product with good physicochemical properties as well as higher organic matter degradation and higher gas production. Moreover, final compost obtained showed good maturity levels and can be used for agricultural applications.

  10. NCSX Vacuum Vessel Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Viola, M. E.; Brown, T.; Heitzenroeder, P.; Malinowski, F.; Reiersen, W.; Sutton, L.; Goranson, P.; Nelson, B.; Cole, M.; Manuel, M.; McCorkle, D.

    2005-10-07

    The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120º vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1" of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120º vessel segments are formed by welding two 60º segments together. Each 60º segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8" (20.3 cm) wide spacer "spool pieces." The vessel must have a total leak rate less than 5 X 10-6 t-l/s, magnetic permeability less than 1.02μ, and its contours must be within 0.188" (4.76 mm). It is scheduled for completion in January 2006.

  11. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Science.gov (United States)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  12. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels; Evaluacion de los defectos inducidos por la radiacion neutronica en los aceros de vasijas

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Jimenez, J.

    1978-07-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs.

  13. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: the case of Fe-Cu model alloys

    CERN Document Server

    Subbotina, A V

    2016-01-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. We show that the obtained results are in a good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  14. Revisiting the reactor pressure vessel for long-time operation; Revisitando la vasija a presion del reactor para largos tiempos de operacion

    Energy Technology Data Exchange (ETDEWEB)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-07-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIFFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  15. Mechanical analysis and reasonable design for Ti-Al alloy liner wound with carbon fiber resin composite high pressure vessel

    Institute of Scientific and Technical Information of China (English)

    Chuan-xiang ZHENG; Fan YANG; Ai-shi ZHU

    2009-01-01

    To consider the internal pressure loaded by both the cylindrical Ti-AI alloy liner and the carbon fiber resin composite (CFRC) wound layers, two models are built. The first one is a cylinder loaded with the internal pressure in the hoop direction only. In this model, the total hoop direction load is distributed over all layers under the internal pressure. The second one is a cylinder loaded with the internal pressure in the axial direction only. In this model, the total axial load is distributed over all cylinders under the internal pressure. Taking the boundary conditions of the continuous displacement between layers into account, a group of equations are built. From these equations, we get the solutions of stresses in both hoop direction and axial direction loaded by every layer under internal pressures. After the stresses are obtained, a reasonable design can be done. An example is given in the final section of this study.

  16. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  17. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  18. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  19. Comparison of MELCOR modeling techniques and effects of vessel water injection on a low-pressure, short-term, station blackout at the Grand Gulf Nuclear Station

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    1995-06-01

    A fully qualified, best-estimate MELCOR deck has been prepared for the Grand Gulf Nuclear Station and has been run using MELCOR 1.8.3 (1.8 PN) for a low-pressure, short-term, station blackout severe accident. The same severe accident sequence has been run with the same MELCOR version for the same plant using the deck prepared during the NUREG-1150 study. A third run was also completed with the best-estimate deck but without the Lower Plenum Debris Bed (BH) Package to model the lower plenum. The results from the three runs have been compared, and substantial differences have been found. The timing of important events is shorter, and the calculated source terms are in most cases larger for the NUREG-1150 deck results. However, some of the source terms calculated by the NUREG-1150 deck are not conservative when compared to the best-estimate deck results. These results identified some deficiencies in the NUREG-1150 model of the Grand Gulf Nuclear Station. Injection recovery sequences have also been simulated by injecting water into the vessel after core relocation started. This marks the first use of the new BH Package of MELCOR to investigate the effects of water addition to a lower plenum debris bed. The calculated results indicate that vessel failure can be prevented by injecting water at a sufficiently early stage. No pressure spikes in the vessel were predicted during the water injection. The MELCOR code has proven to be a useful tool for severe accident management strategies.

  20. Investigation of the Impact of ENDF/B-VI Cross Sections on the H.B. Robinson-2 Pressure-Vessel Flux Prediction

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I

    1999-06-01

    This report discusses the impact of the change from the SAILOR cross-section library, based on the ENDF/B-IV data, to the BUGLE-96 cross-section library, based on the ENDF/B-VI data, on the neutron flux prediction in the H. B. Robinson-2 pressure vessel, in the surveillance capsule, and in the cavity. The fast flux (E > 1 MeV) from the transport calculations with the BUGLE-96 library is {approximately}6% higher in the surveillance capsule and at the PV inner wall, and {approximately}25% higher in the reactor cavity than the flux from the transport calculations with the SAILOR library. These changes result from the combined effect of the changes in the cross sections, which cause significant increases in the calculated fluxes, and much smaller decreases in the fast fluxes due to the changes in the fission spectra. The increase in the calculated fast flux due to the changes in the cross sections only is {approximately}9% in the capsule and at the pressure vessel (PV) wall, and {approximately}30% in the cavity. The changes in the fission spectra lead to decreases in the order of {approximately}3-4% in calculated fast fluxes.

  1. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  2. Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.

  3. 10 MW高温气冷堆反应堆压力容器的出厂水压试验%Hydraulic Pressure Test of Pressure Vessel of 10 MW High Temperature Gas-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    刘俊杰; 张征明; 何树延; 王金海

    2001-01-01

    The hydraulic pressure test of 10MW Hight Temperature Gas-cooled Reactorc(HTR-10) pressure vessel was successfully performed according to the requirement of the section NB-6200, ASME Ⅲ code. The test requirement, the test results and the test evaluations are described in detail. The test tension was effectively and rationally done through an hydraulic tensionor, which was developed at institue of nuclear energy technology of Tsinghua University. The strain and deformation of the HTR-10 pressure vessel were also measured.%根据ASME规范第Ⅲ卷NB-6200节的规定,对10MW高温气冷堆压力容器的水压试验要求、试验过程,试验结果及评价进行了叙述。用清华大学核能技术设计研究院研制的液压张拉机对主螺栓实施了合理及有效的张拉,对压力容器进行了应变和变形测量,取得了反应堆压力容器水压试验的圆满成功。

  4. 反应堆退役压力容器放射性活度估算方法%Method for Estimation of Activity in Decommissioned Nuclear Reactor Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    郭武仁; 林晓玲; 郑宁宁

    2011-01-01

    The theoretical calculation and experimental measurement methods for estimation of activity in the decommissioned nuclear reactor pressure vessel were introduced. The physical estimation model was described,and Monte Carlo compute code and 0RIGEN2 code were recommended to be employed for the calculation of the neutron flux and activity in the reactor pressure vessel. Two methods commonly used for determining the activity in the reactor pressure vessel were introduced in detail,I.e.,sampling from the reactor pressure vessel and from the irradiation tube. The neutron flux profile was established to predict the activity in the reactor pressure vessel.%介绍了反应堆退役压力容器放射性活度估算的理论计算和实验测定方法.描述了物理估算模型,推荐采用蒙特卡罗程序和ORIGEN2程序分别计算中子通量密度和放射性活度.对确定压力容器的放射性活度时经常使用的两种方法(压力容器直接取样分析和对辐照监督管取样分析)做了详细介绍.建立了推算压力容器的放射性活度中子通量密度比例曲线.

  5. Rapid South Atlantic spreading changes and coeval vertical motion in surrounding continents: Evidence for temporal changes of pressure-driven upper mantle flow

    DEFF Research Database (Denmark)

    Colli, L.; Stotz, Ingo Leonardo; Bunge, H.P.

    2014-01-01

    the rapid spreading rate changes, on order 10 million years, require significant decoupling of regional plate motion from the large-scale mantle buoyancy distribution through a mechanically weak asthenosphere. Andean topographic growth in late Miocene can explain the most recent South Atlantic spreading...... velocity reduction, arising from increased plate boundary forcing associated with the newly elevated topography. But this mechanism is unlikely to explain the Late Cretaceous/Tertiary spreading variations, as changes in Andean paleoelevation at the time are small. We propose an unsteady pressure...

  6. Three dimensional parameterization modeling method for pressure vessel head%压力容器封头参数化三维建模方法

    Institute of Scientific and Technical Information of China (English)

    张义顺; 梁盈; 刘海波

    2011-01-01

    为了满足压力容器设计行业的需要,提高压力容器封头、简体、法兰及弯头等组件的设计效率,采用内嵌在AutoCAD内部的VBA编程工具建立UCS,编写相关函数,增加插入点功能并融合空间坐标系转换技术,利用布尔运算、旋转等方法实现了封头三维建模.利用该参数化建模方法,设计人员只需输入封头的相关数据,便可在任意视角下、任意捕捉点处得到符合要求的三维封头模型.此外,结合AutoCAD三维造型技术,配合VBA编程,可对压力容器其他组件如法兰、弯头等进行参数化建模,这对于开发设计相关行业AutoCAD三维插件具有一定的指导意义.%To meet the need of pressure vessel design industry and raise the design efficiency of such components as pressure vessel head, cylinder, flange and elbow, UCS was established with VBA programming tool embedded in AutoCAD software and the relative functions were compiled. Through adding the insertion point function and combing the transform technology of spatial coordinate system, three-dimensional modeling for the head was realized with such methods as Boolean operation and rotation. With the proposed parameterization modeling method, 3D head model to meet the requirement can be obtained at arbitrary visual angle and target point only through entering the relevant data of the head by designer. Besides, in combination with AutoCAD 3D modeling technology and VBA programming, the parameterization modeling for other pressure vessel components such as flange and elbow can be performed. The present research can provide a certain reference for exploiting and designing the AutoCAD 3D plug-in in relevant industries.

  7. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    The role of the research vessels as a tool for marine research and exploration is very important. Technical requirements of a suitable vessel and the laboratories needed on board are discussed. The history and the research work carried out...

  8. In-Situ Nondestructive Evaluation of Kevlar(Registered Trademark)and Carbon Fiber Reinforced Composite Micromechanics for Improved Composite Overwrapped Pressure Vessel Health Monitoring

    Science.gov (United States)

    Waller, Jess; Saulsberry, Regor

    2012-01-01

    NASA has been faced with recertification and life extension issues for epoxy-impregnated Kevlar 49 (K/Ep) and carbon (C/Ep) composite overwrapped pressure vessels (COPVs) used in various systems on the Space Shuttle and International Space Station, respectively. Each COPV has varying criticality, damage and repair histories, time at pressure, and pressure cycles. COPVs are of particular concern due to the insidious and catastrophic burst-before-leak failure mode caused by stress rupture (SR) of the composite overwrap. SR life has been defined [1] as the minimum time during which the composite maintains structural integrity considering the combined effects of stress level(s), time at stress level(s), and associated environment. SR has none of the features of predictability associated with metal pressure vessels, such as crack geometry, growth rate and size, or other features that lend themselves to nondestructive evaluation (NDE). In essence, the variability or surprise factor associated with SR cannot be eliminated. C/Ep COPVs are also susceptible to impact damage that can lead to reduced burst pressure even when the amount of damage to the COPV is below the visual detection threshold [2], thus necessitating implementation of a mechanical damage control plan [1]. Last, COPVs can also fail prematurely due to material or design noncompliance. In each case (SR, impact or noncompliance), out-of-family behavior is expected leading to a higher probability of failure at a given stress, hence, greater uncertainty in performance. For these reasons, NASA has been actively engaged in research to develop NDE methods that can be used during post-manufacture qualification, in-service inspection, and in-situ structural health monitoring. Acoustic emission (AE) is one of the more promising NDE techniques for detecting and monitoring, in real-time, the strain energy release and corresponding stress-wave propagation produced by actively growing flaws and defects in composite

  9. Determination of an adequate perfusion pressure for continuous dual vessel hypothermic machine perfusion of the rat liver

    NARCIS (Netherlands)

    't Hart, Nils A.; der van Plaats, Arjan; Leuvenink, Henri G. D.; van Goor, Harry; Wiersema-Buist, Janneke; Verkerke, Gijsbertus J.; Rakhorst, Gerhard; Ploeg, Rutger J.

    Hypothermic machine perfusion (HMP) provides better protection against ischemic damage of the kidney compared to cold-storage. The required perfusion pressures needed for optimal HMP of the liver are, however, unknown. Rat livers were preserved in University of Wisconsin organ preservation solution

  10. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Science.gov (United States)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  11. Study of atomic clusters in neutron irradiated reactor pressure vessel surveillance samples by extended X-ray absorption fine structure spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Cammelli, S. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)], E-mail: Sebastiano.cammelli@psi.ch; Degueldre, C.; Kuri, G.; Bertsch, J. [LWV, NES, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Luetzenkirchen-Hecht, D.; Frahm, R. [Fachbereich C - Physik, Bergische Universitaet Wuppertal, Gauss-Str. 20, 42097 Wuppertal (Germany)

    2009-03-31

    Copper and nickel impurities in nuclear reactor pressure vessel (RPV) steel can form nano-clusters, which have a strong impact on the ductile-brittle transition temperature of the material. Thus, for control purposes and simulation of long irradiation times, surveillance samples are submitted to enhanced neutron irradiation. In this work, surveillance samples from a Swiss nuclear power plant were investigated by extended X-ray absorption fine structure spectroscopy (EXAFS). The density of Cu and Ni atoms determined in the first and second shells around the absorber is affected by the irradiation and temperature. The comparison of the EXAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the EXAFS analysis reveals local irradiation damage in the form of vacancy fractions, which can be determined with a precision of {approx}5%. There are indications that the formation of Cu and Ni clusters differs significantly.

  12. Pressure vessel fracture studies pertaining to a PWR LOCA-ECC thermal shock: experiments TSE-3 and TSE-4 and update of TSE-1 and TSE-2 analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Bolt, S.E.

    1977-11-04

    The LOCA-ECC Thermal Shock Program was established to investigate the potential for flaw propagation in pressurized-water reactor (PWR) vessels during injection of emergency core coolant following a loss-of-coolant accident. Studies thus far have included fracture mechanics analyses of typical PWRs, the design and construction of a thermal shock test facility, determination of material properties for test specimens, and four thermal shock experiments with 0.53-m-OD (21-in.) by 0.15-m-wall (6-in.) cylindrical test specimens. In the first experiment, initiation was not expected and did not occur, although there was a small amount of subcritical crack growth. In the second experiment, initiation of a semicircular flaw took place as expected; the final length along the surface was about four times the initial length, but there was no radial growth. The third and fourth experiments were similar, and the long axial flaw initiated in good agreement with predictions.

  13. The investigation and analysis about the defects detected in the piping and the pressure vessel of the Kori Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Eun Soo; Park, Ik Geon; Lee, Jeong Soon; Kim, Hyeon Ju; Eom, Byeong Gook; Kim, Hyeon Mook; Cho, Dong Soo [Seoul National Univ. of Technology, Seoul (Korea, Republic of)

    1997-07-15

    The objective of this report os to analysis and investigate the defects detected in the pipe and the pressure vessel of the Kori Nuclear Power Plants during PSI/ISI. In this report, an intelligent database program on windows 95, computer operating system has been built for the defects in the Kori nuclear power plant during PSI/ISI. An intelligent data bases program has been constructed for the effective management of NDE(Nondestructive Evaluation) data carried out the Kori nuclear power plant. Data bases program can be applied to statistical analysis and investigation of the defect data detected during PSI/ISI under fully compatible with windows 95. It is also possible to investigate the NDE data inspected repetitively and the contents of treatment about them.

  14. Fracture toughness evaluation of 20MnMoNi55 pressure vessel steel in the ductile to brittle transition regime: Experiment & numerical simulations

    Science.gov (United States)

    Gopalan, Avinash; Samal, M. K.; Chakravartty, J. K.

    2015-10-01

    In this work, fracture behaviour of 20MnMoNi55 reactor pressure vessel (RPV) steel in the ductile to brittle transition regime (DBTT) is characterised. Compact tension (CT) and single edged notched bend (SENB) specimens of two different sizes were tested in the DBTT regime. Reference temperature 'T0' was evaluated according to the ASTM E1921 standard. The effect of size and geometry on the T0 was studied and T0 was found to be lower for SENB geometry. In order to understand the fracture behaviour numerically, finite element (FE) simulations were performed using Beremin's model for cleavage and Rousselier's model for ductile failure mechanisms. The simulated fracture behaviour was found to be in good agreement with the experiment.

  15. 等极孔球形压力容器平面缠绕规律%Flat Winding Principle About Equal Polar Hole Spherical Pressure Vessel

    Institute of Scientific and Technical Information of China (English)

    王洪运; 马国峰; 赵亮; 费春东

    2012-01-01

    Based on the theory of space analytic geometry and kinematics,fiber-touched point' s locus equation of flat winding about FRP spherical pressure vessel is established. The functions of fiber-touched point' s space location, and velocity and acceleration of equal polar hole spherical pressure vessel' s flat winding at any time are deduced. Velocity in various directions and cosine value of angular separation of the fiber-touched point were solved by the analysis of doff point's winding principle. Finally,winding locus is simulated with various initial angles by software of Matlab,the effectiveness of the proposed approach is validated by the comparison results of numerical simulation and actual fiber winding style.%基于空间解析几何理论及运动学分析方法,建立复合材料球形压力容器平面缠绕落纱点轨迹运动方程,推导出等极孔球形容器表面缠绕纤维落纱点空间位置、速度以及加速度的求解函数;通过对落纱点运动规律的分析,解出空间落纱点各方向运动速度的大小及其各方向夹角余弦值.最后利用Matlab软件对不同的丝嘴运动平面倾斜角的缠绕轨迹进行模拟,通过该计算方法模拟的轨迹结果与实际球形容器缠绕线型相符合,验证了该算法的有效性.

  16. Fracture mechanics characterisation of the beltline welding seam of the decommissioned WWER-440 reactor pressure vessel of nuclear power plant Greifswald Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, Hans-Werner, E-mail: H.W.Viehrig@hzdr.de [Helmholz Zentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany); Altstadt, Eberhard; Houska, Mario [Helmholz Zentrum Dresden-Rossendorf, P.O. Box 510119, 01314 Dresden (Germany); Valo, Matti [VTT Manufacturing Technology, P.O. Box 17042, 02044 VTT (Finland)

    2012-01-15

    The paper presents data measured for trepans sampled from the decommissioned WWER-440 reactor pressure vessel of the NPP Greifswald Unit 4. The main focus of this work is on fracture toughness characterisation according to test standard ASTM E1921. Large variations in the evaluated reference temperature values, T{sub 0}, across the wall of the multilayer beltline welding seam were observed. Generally, the ductile-to-brittle transition temperature shift predicted by the Russian code for the present content of deleterious elements P and Cu and the accumulated neutron fluences lies within the amount of the scatter of the measured T{sub 0} values. Metallographic investigations show that the T{sub 0} values measured with T-S oriented Charpy size SE(B) specimens from different thickness locations of the multilayer welding seams strongly depend on the microstructure at the specimen crack tip, and, consequently, on the initial structure of the multilayer welding seam. The RPV integrity is discussed, taking into account a pressurised thermal shock scenario. - Highlights: Black-Right-Pointing-Pointer The paper presents data of samples from a decommissioned reactor pressure vessel. Black-Right-Pointing-Pointer The main focus is on fracture toughness characterisation of the beltline weld seam. Black-Right-Pointing-Pointer Large variation in the evaluated reference temperatures T{sub 0} was observed. Black-Right-Pointing-Pointer T{sub 0} values strongly depend on the microstructure at the specimen crack tip. Black-Right-Pointing-Pointer RPV integrity is discussed, taking into account a pressurised thermal shock scenario.

  17. The Assessment and Validation of Mini-Compact Tension Test Specimen Geometry and Progress in Establishing Technique for Fracture Toughness Master Curves for Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only a small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. A small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes a variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels (RPVs) since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Assessment and validation of mini-CT specimen geometry has been performed on previously well characterized HSST Plate 13B, an A533B class 1 steel. It was shown that the fracture toughness transition temperature measured by these Mini-CT specimens is within the range of To values that were derived from various large fracture toughness specimens. Moreover, the scatter of the fracture toughness values measured by Mini-CT specimens perfectly follows the Weibull distribution function providing additional proof for validation of this geometry for the Master Curve evaluation of rector pressure vessel steels. Moreover, the International collaborative program has been developed to extend the assessment and validation efforts to irradiated weld metal. The program is underway and involves ORNL, CRIEPI, and EPRI.

  18. Health Monitoring of Composite Overwrapped Pressure Vessels (COPVs) Using Meandering Winding Magnetometer ((MWM(Registered Trademark)) Eddy Current Sensors

    Science.gov (United States)

    Russell, Rick; Grundy, David; Jablonski, David; Martin, Christopher; Washabaugh, Andrew; Goldfine, Neil

    2011-01-01

    There are 3 mechanisms that affect the life of a COPV are: a) The age life of the overwrap; b) Cyclic fatigue of the metallic liner; c) Stress Rupture life. The first two mechanisms are understood through test and analysis. A COPV Stress Rupture is a sudden and catastrophic failure of the overwrap while holding at a stress level below the ultimate strength for an extended time. Currently there is no simple, deterministic method of determining the stress rupture life of a COPV, nor a screening technique to determine if a particular COPV is close to the time of a stress rupture failure. Conclusions: Demonstrated a correlation between MWM response and pressure or strain. Demonstrated the ability to monitor stress in COPV at different orientations and depths. FA41 provides best correlation with bottle pressure or stress.

  19. Experimental Analysis of the Influence of Hydrostatic Stress on the Behaviour of an Adhesive Using a Pressure Vessel

    OpenAIRE

    Cognard, J. Y.; Creac' Hcadec, R; da Silva, L. F. M.; Teixeira, F. G.; Davies, Peter; Peleau, Michel

    2011-01-01

    The modelling of the non-linear behaviour of ductile adhesives requires a large experimental database in order to represent accurately the strains which are strongly dependent on the tensile-shear loading combination. Various pressure-dependent constitutive models can be found in the literature, but only a few experimental results are available, for instance, to represent accurately the initial yield surface taking into account the two stress invariants, hydrostatic stress, and von Mises equi...

  20. The surface chemistry resulting from low-pressure plasma treatment of polystyrene: The effect of residual vessel bound oxygen

    Science.gov (United States)

    Dhayal, Marshal; Alexander, Morgan R.; Bradley, James W.

    2006-09-01

    The surface chemistry of plasma treated polystyrene samples has been studied in a specially designed low-pressure argon discharge system incorporating in situ XPS analysis. By using an electrostatic grid biasing technique, the plasma source can also be used in a mode preventing ion interactions with the sample. The system, which utilizes a vacuum transfer chamber between plasma and XPS analysis has allowed us to differentiate between the level of oxygen incorporated at the polystyrene surface from residual gas during treatment and that from the exposure of the treated sample to the laboratory atmosphere. Using typical base pressures of about 5 × 10 -3 Pa (4 × 10 -5 Torr) the XPS results show that significant oxygen surface incorporation resulted from oxygen containing species in the plasma itself (i.e. water vapour with 2 × 10 -3 Pa partial pressure). The surface concentration of O was measured at 7.6 at.%. Subsequent atmospheric exposure of the treated samples resulted in only a small increase (of 0.6 at.%) in oxygen incorporation in the form of acid anhydride functionalities. XPS measurements of PS samples exposed to plasmas with no ion-surface component (i.e. exposure from VUV, UV and excited neutral species only) showed no appreciable change in oxygen incorporation compared to those with low-energy ion bombardment from the plasma (free radical sites in this discharge regime.