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Sample records for nuclear pressure vessels

  1. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  2. Cover for a nuclear reactor pressure vessel

    SciTech Connect

    Gross, H.

    1980-03-11

    A pressure vessel, containment or burst shield for a nuclear reactor has a substantially circular cover surmounting the cylindrical part (Shell) of the vessel and is preferably comprised of a plurality of circular or polylateral segments arranged concentrically and stressed inwardly by annular prestressing means. At least the outer polylateral segments and preferably all of the circular segments are provided on the upper surface with upwardly open circular grooves receiving the prestressing arrangement. The latter can comprise an outwardly open channel-shaped (U-section) supporting member receiving the stressing cables and means for transferring the radial stress of the annular stressing arrangement to the ring segment. The latter means may be wedges inserted between the support and a wall of the groove after the stressing arrangement has been placed under stress, E.G. By hydraulic means for spreading the annular stressing arrangement.

  3. Structural integrity of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    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.

  4. High pressure, high-temperature vessel, especially for nuclear reactors

    SciTech Connect

    Mitterbacher, P.; Schoning, J.; Schwiers, H.G.

    1980-09-23

    A pressure vessel susceptible to high temperatures, especially for containment of a nuclear-reactor core, is constituted of a cylindrical shell from a cast material such as cast steel, cast iron or concrete, and is prestressed by vertical cables which extend parallel to generatrices of the shell. Peripheral (Circumferential) prestressing cables are provided around the shell which can be externally insulated. The peripheral tensioning cables are exposed externally of the insulation material and bear upon the shell of the vessel with heatresistant elements of high compressive strength which extend through the external insulation.

  5. Continuous Cooling Transformations in Nuclear Pressure Vessel Steels

    NASA Astrophysics Data System (ADS)

    Pous-Romero, Hector; Bhadeshia, Harry K. D. H.

    2014-10-01

    A class of low-alloy steels often referred to as SA508 represent key materials for the manufacture of nuclear reactor pressure vessels. The alloys have good properties, but the scatter in properties is of prime interest in safe design. Such scatter can arise from microstructural variations but most studies conclude that large components made from such steels are, following heat treatment, fully bainitic. In the present work, we demonstrate with the help of a variety of experimental techniques that the microstructures of three SA508 Gr.3 alloys are far from homogeneous when considered in the context of the cooling rates encountered in practice. In particular, allotriomorphic ferrite that is expected to lead to a deterioration in toughness, is found in the microstructure for realistic combinations of austenite grain size and the cooling rate combination. Parameters are established to identify the domains in which SA508 Gr.3 steels transform only into the fine bainitic microstructures.

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

    NASA Astrophysics Data System (ADS)

    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.

  7. Damage dosimetry and embrittlement monitoring of nuclear pressure vessels in real time by magnetic properties measurement

    SciTech Connect

    Stubbins, J.F.; Ougouag, A.M.; Williams, J.G.

    1992-07-01

    The objective of this project is to develop a technique for real-time monitoring of neutron dose and of the onset and progression of embrittlement in operating nuclear pressure vessels. The technique relies on the measurement of magnetic properties of steel and other magnetic materials which are extremely sensitive to radiation-induced properties changes. The approach being developed here is innovative and unique. It promises to be readily applicable to all existing and planned reactor structures. The significance of this program is that it addresses a major concern in the operation of existing nuclear pressure vessels. The development of microscopic defect clusters during irradiation in the nuclear pressure vessel beltline region leads to an increase in material yield strength and a concomitant decrease in ductility, or ability to absorb energy in fracture (i.e. fracture toughness). This decrease in fracture toughness is alarming since it may impair the ability of the pressure vessel to resist fracture during unusual loading situations.

  8. Non-invasive liquid level and density gauge for nuclear power reactor pressure vessels

    SciTech Connect

    Baratta, A.J.; Jester, W.A.; Kenney, E.S.; Mc Master, I.B.; Schultz, M.A.

    1987-01-27

    A method is described of non-invasively determining the liquid coolant level and density in a nuclear power reactor pressure vessel comprising the steps: positioning at least three neutron detector fission chambers externally of the reactor pressure vessel at multiple spaced positions along the side of the fuel core. One of the neutron detectors is positioned at the side near the bottom of the fuel core. The multiple spaced positions along the side remove any ambiguity as to whether the liquid level is decreasing or increasing: shielding the neutron detector fission chamber from thermal neutrons to avoid the noise associated therewith, and eliminating the effects of gamma radiation from the detected signals; monitoring the detected neutron level signals to determine to coolant liquid level and density in the nuclear power reactor pressure vessel.

  9. Evaluating the safety of aging nuclear reactor pressure vessels

    SciTech Connect

    Pennell, W.E.

    1996-05-01

    Regulatory requirements limit the permissible accumulation of irradiation damage in RPV material such that adequate fracture prevention margins are maintained throughout the licensed operating period of a nuclear plant. Experience with application of those requirements has identified a number of areas where they could be further refined to eliminate excess conservatism. Research is ongoin to provide the data required to support refinement of the regulatory requirements. Research programs are investigating theeffects of local brittle zones, shallow flaws, biaxial loading, and stainless steel cladding. Preliminary results from this research indicate a potential for beneficial changes in the P-T curve and PTS analysis rules.

  10. Reactor pressure vessel structural integrity research in the US Nuclear Regulatory Commission HSST and HSSI Programs

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1994-02-01

    This report discusses development on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels containing flaws. Fracture mechanics tests on reactor pressure vessel steel have shown that local brittle zones do not significantly degrade the material fracture toughness, constraint relaxation at the crack tip of shallow surface flaws results in increased fracture toughness, and biaxial loading reduces but does not eliminate the shallow-flaw fracture toughness elevation. Experimental irradiation investigations have shown that the irradiation-induced shift in Charpy V-notch versus temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement and 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.

  11. Fast neutron fluence of yonggwang nuclear unit 1 reactor pressure vessel

    SciTech Connect

    Yoo, C.; Km, B.; Chang, K.; Leeand, S.; Park, J.

    2006-07-01

    The Code of Federal Regulations, Title 10, Part 50, Appendix H, requires that the neutron dosimetry be present to monitor the reactor vessel throughout plant life. The Ex-Vessel Neutron Dosimetry System has been installed for Yonggwang Nuclear Unit 1 after complete withdrawal of all six in-vessel surveillance capsules. This system has been installed in the reactor cavity annulus in order to measure the fast neutron spectrum coming out through the reactor pressure vessel. Cycle specific neutron transport calculations were performed to obtain the energy dependent neutron flux throughout the reactor geometry including dosimetry positions. Comparisons between calculations and measurements were performed for the reaction rates of each dosimetry sensors and results show good agreements. (authors)

  12. Magnetic hysteresis properties of neutron-irradiated VVER440-type nuclear reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.; Horváth, M.

    2012-11-01

    The development of non-destructive evaluation methods for irradiation embrittlement in nuclear reactor pressure vessel steels has a key role for safe and long-term operation of nuclear power plants. In this study, we have investigated the effect of neutron irradiation on base and weld metals of Russian VVER440-type reactor pressure vessel steels by measurements of magnetic minor hysteresis loops. A minor-loop coefficient, which is obtained from a scaling power-law relation of minor-loop parameters and is a sensitive indicator of internal stress, is found to change with neutron fluence for both metals. While the coefficient for base metal exhibits a local maximum at low fluence and a subsequent slow decrease, that for weld metal monotonically decreases with fluence. The observed results are explained by competing mechanisms of nanoscale defect formation and recovery, among which the latter process plays a dominant role for magnetic property changes in weld metal due to its ferritic microstructure.

  13. Prevention of non-ductile fracture in 6061-T6 aluminum nuclear pressure vessels

    SciTech Connect

    Yahr, G.T.

    1995-06-01

    The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Committee has approved rules for the use of 6061-T6 and 6061-T651 aluminum for the construction of Class 1 welded nuclear pressure vessels for temperatures not exceeding 149 C (300 F). Nuclear Code Case N-519 allows the use of this aluminum in the construction of low temperature research reactors such as the Advanced Neutron Source. The rules for protection against non-ductile fracture are discussed. The basis for a value of 25.3 MPa {radical}m (23 ksi {radical}in.) for the critical or reference stress intensity factor for use in the fracture analysis is presented. Requirements for consideration of the effects of neutron irradiation on the fracture toughness are discussed.

  14. Ab initio simulation of radiation damage in nuclear reactor pressure vessel materials

    NASA Astrophysics Data System (ADS)

    Watts, Daniel; Finkenstadt, Daniel

    2012-02-01

    Using Kinetic Monte Carlo we developed a code to study point defect hopping in BCC metallic alloys using energetics and attempt frequencies calculated using VASP, an electronic structure software package. Our code provides a way of simulating the effects of neutron radiation on potential reactor materials. Specifically we will compare the Molybdenum-Chromium alloy system to steel alloys for use in nuclear reactor pressure vessels.

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

    NASA Astrophysics Data System (ADS)

    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.

  16. Development of temper-bead technique applied to dissimilar welded joints of nuclear pressure vessels

    SciTech Connect

    Higuchi, Makoto; Umemoto, Tadahiro; Matsusita, Akitake; Shiraiwa, Takanori

    1996-06-01

    When nuclear pressure vessels made of low-alloy steel (P-3 Group 3) need repair or modification, technical standards for welding of electrical structures should be applied, and then postweld heat treatment (PWHT) should be done. However, cases in which PWHT is impractical are theoretically possible due to a variety of restrictions. To deal with such a problem, there is a regulation for repair weld technique, without PWHT, in accordance with ASME B and PV Code. This method is called temper-bead technique, which gives the weldments sufficient toughness by tempering the hardened zone of the heat-affected zone on the first layer of the base metal using the heat of the following weld beads. Because there is no regulation in Japan covering this method, a procedure is required to perform it under a special license, after a verification test has been passed. An attempt has been made to develop a method, on the supposition that the temper-bead technique is adopted for replacement of what is called dissimilar welded joints, so that a nickel base alloy is buildup welded at the tip of the nozzle of the low-alloy steel pressure vessel, and a stainless steel pipe is butt welded.

  17. Assessment of Radiation Embrittlement in Nuclear Reactor Pressure Vessel Surrogate Materials

    NASA Astrophysics Data System (ADS)

    Balzar, Davor

    2010-10-01

    The radiation-enhanced formation of small (1-2 nm) copper-rich precipitates (CRPs) is critical for the occurrence of embrittlement in nuclear-reactor pressure vessels. Small CRPs are coherent with the bcc matrix, which causes local matrix strain and interaction with the dislocation strain fields, thus impeding dislocation mobility. As CRPs grow, there is a critical size at which a phase transformation occurs, whereby the CRPs are no longer coherent with the matrix, and the strain is relieved. Diffraction-line-broadening analysis (DLBA) and small-angle neutron scattering (SANS) were used to characterize the precipitate formation in surrogate ferritic reactor-pressure vessel steels. The materials were aged for different times at elevated temperature to produce a series of specimens with different degrees of copper precipitation. SANS measurements showed that the precipitate size distribution broadens and shifts toward larger sizes as a function of ageing time. Mechanical hardness showed an increase with ageing time, followed by a decrease, which can be associated with the reduction in the number density as well as the loss of coherency at larger sizes. Inhomogeneous strain correlated with mechanical hardness.

  18. Development of a shallow-flaw fracture assessment methodology for nuclear reactor pressure vessels

    SciTech Connect

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

    1996-06-01

    Shallow-flaw fracture technology is being developed within the Heavy-Section Steel Technology (HSST) Program for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVs) containing postulated shallow flaws. Cleavage fracture in shallow-flaw cruciform beam specimens tested under biaxial loading at temperatures in the lower transition temperature range was shown to be strain-controlled. A strain-based dual-parameter fracture toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture. A probabilistic fracture mechanics (PFM) model that includes both the properties of the inner-surface stainless-steel cladding and a biaxial shallow-flaw fracture toughness correlation gave a reduction in probability of cleavage initiation of more than two orders of magnitude from an ASME-based reference case.

  19. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, R.C.; Upton, H.A.

    1994-10-04

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

  20. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

  1. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept

    NASA Technical Reports Server (NTRS)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.

    1973-01-01

    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  2. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    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

  3. Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

    NASA Astrophysics Data System (ADS)

    Tomáš, I.; Vértesy, G.; Gillemot, F.; Székely, R.

    2013-01-01

    Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 1019-11.9 × 1019 n/cm2. Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile-brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.

  4. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  5. Radiation embrittlement of nuclear reactor pressure vessel steels: An international review (Fourth Volume)

    SciTech Connect

    Steele, L.E.

    1993-12-01

    The technical content is highly focused on the title subject, which is crucial to the continued operating safety of commercial nuclear electric power generating plants, as it treats the phenomenon of neutron embrittlement of the primary containment vessel of the nuclear reactor power source. Integrity of this nuclear reactor component is a primary goal of all the specialists who have participated in this series of four international meetings. These international meetings and the publication arising from them offer a progressive series of volumes that provide a valuable technical resource to nuclear power plant operators, national regulatory specialists, and researchers in this area of nuclear safety. The progressive nature of these publications is particularly valuable in teaching scientific and technical developments on what has become one of the most critical elements in reactor safety analysis with the aging of nuclear power reactors. The progress of research and vessel surveillance for neutron embrittlement reflects the aging of nuclear power reactors and, therefore, the attendant interest in assuring safe life attainment for this crucial element of electric power generation, as the authors approach the close of the Twentieth century. Separate abstracts were prepared for 31 papers of this book.

  6. Dual shell pressure balanced vessel

    DOEpatents

    Fassbender, Alexander G.

    1992-01-01

    A dual-wall pressure balanced vessel for processing high viscosity slurries at high temperatures and pressures having an outer pressure vessel and an inner vessel with an annular space between the vessels pressurized at a pressure slightly less than or equivalent to the pressure within the inner vessel.

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

    ERIC Educational Resources Information Center

    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)…

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

    ERIC Educational Resources Information Center

    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)…

  9. Pressure vessel flex joint

    NASA Technical Reports Server (NTRS)

    Kahn, Jon B. (Inventor)

    1992-01-01

    An airtight, flexible joint is disclosed for the interfacing of two pressure vessels such as between the Space Station docking tunnel and the Space Shuttle Orbiter bulkhead adapter. The joint provides for flexibility while still retaining a structural link between the two vessels required due to the loading created by the internal/external pressure differential. The joint design provides for limiting the axial load carried across the joint to a specific value, a function returned in the Orbiter/Station tunnel interface. The flex joint comprises a floating structural segment which is permanently attached to one of the pressure vessels through the use of an inflatable seal. The geometric configuration of the joint causes the tension between the vessels created by the internal gas pressure to compress the inflatable seal. The inflation pressure of the seal is kept at a value above the internal/external pressure differential of the vessels in order to maintain a controlled distance between the floating segment and pressure vessel. The inflatable seal consists of either a hollow torus-shaped flexible bladder or two rolling convoluted diaphragm seals which may be reinforced by a system of straps or fabric anchored to the hard structures. The joint acts as a flexible link to allow both angular motion and lateral displacement while it still contains the internal pressure and holds the axial tension between the vessels.

  10. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect

    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

  11. Pressurized Vessel Slurry Pumping

    SciTech Connect

    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.

  12. Sapphire tube pressure vessel

    DOEpatents

    Outwater, John O.

    2000-01-01

    A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

  13. Sapphire tube pressure vessel

    SciTech Connect

    Outwater, J.O.

    2000-05-23

    A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

  14. Pressure vessel bottle mount

    NASA Technical Reports Server (NTRS)

    Wingett, Paul (Inventor)

    2001-01-01

    A mounting assembly for mounting a composite pressure vessel to a vehicle includes a saddle having a curved surface extending between two pillars for receiving the vessel. The saddle also has flanged portions which can be bolted to the vehicle. Each of the pillars has hole in which is mounted the shaft portion of an attachment member. A resilient member is disposed between each of the shaft portions and the holes and loaded by a tightening nut. External to the holes, each of the attachment members has a head portion to which a steel band is attached. The steel band circumscribes the vessel and translates the load on the springs into a clamping force on the vessel. As the vessel expands and contracts, the resilient members expand and contract so that the clamping force applied by the band to the vessel remains constant.

  15. Attachment Fitting for Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Smeltzer, Stanley S., III (Inventor); Carrigan, Robert W. (Inventor)

    2002-01-01

    This invention provides sealed access to the interior of a pressure vessel and consists of a tube. a collar, redundant seals, and a port. The port allows the seals to be pressurized and seated before the pressure vessel becomes pressurized.

  16. Attachment Fitting for Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Smeltzer, Stanley S., III (Inventor); Carrigan, Robert W. (Inventor)

    2002-01-01

    This invention provides sealed access to the interior of a pressure vessel and consists of a tube. a collar, redundant seals, and a port. The port allows the seals to be pressurized and seated before the pressure vessel becomes pressurized.

  17. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    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.

  18. Improved mechanical properties of A 508 class 3 steel for nuclear pressure vessel through steelmaking

    SciTech Connect

    Kim, J.T.; Kwon, H.K.; Kim, K.C.; Kim, J.M.

    1997-12-31

    The present work is concerned with the steelmaking practices which improve the mechanical properties of the A 508 class 3 steel for reactor pressure vessel. Three kinds of steelmaking practices were applied to manufacture the forged heavy wall shell for reactor pressure vessel, that is, the vacuum carbon deoxidation (VCD), modified VCD containing aluminum and silicon-killing. The segregation of the chemical elements through the thickness was quite small so that the variations of the tensile properties at room temperature were small and the anisotropy of the impact properties was hardly observed regardless of the steelmaking practices. The Charpy V-notch impact properties and the reference nil-ductile transition temperature by drop weight test were significantly improved by the modified VCD and silicon-killing as compared with those of the steel by VCD. Moreover, the plane strain fracture toughness values of the materials by modified VCD and silicon-killing practices was much higher than those of the steel by VCD. These were resulted from the fining of austenite grain size. It was observed that the grain size was below 20 {micro}m (ASTM No. 8.5) when using the modified VCD and silicon-killing, compared to 50 {micro}m (ASTM No. 7.0) when using VCD.

  19. Damage dosimetry and embrittlement monitoring of nuclear pressure vessels in real time by magnetic properties measurement. Technical progress report for year 2, October 1, 1991--September 30, 1992

    SciTech Connect

    Stubbins, J.F.; Ougouag, A.M.; Williams, J.G.

    1992-07-01

    The objective of this project is to develop a technique for real-time monitoring of neutron dose and of the onset and progression of embrittlement in operating nuclear pressure vessels. The technique relies on the measurement of magnetic properties of steel and other magnetic materials which are extremely sensitive to radiation-induced properties changes. The approach being developed here is innovative and unique. It promises to be readily applicable to all existing and planned reactor structures. The significance of this program is that it addresses a major concern in the operation of existing nuclear pressure vessels. The development of microscopic defect clusters during irradiation in the nuclear pressure vessel beltline region leads to an increase in material yield strength and a concomitant decrease in ductility, or ability to absorb energy in fracture (i.e. fracture toughness). This decrease in fracture toughness is alarming since it may impair the ability of the pressure vessel to resist fracture during unusual loading situations.

  20. GOLD PRESSURE VESSEL SEAL

    DOEpatents

    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)

  1. Magnetic properties of a highly neutron-irradiated nuclear reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Kobayashi, S.; Gillemot, F.; Horváth, Á.; Székely, R.

    2012-02-01

    We report results of minor B- H loop measurements on a highly neutron-irradiated A533B-type reactor pressure vessel steel. A minor-loop coefficient, which is a sensitive indicator of internal stress, changes with neutron fluence, but depends on relative orientation to the rolling direction in the low fluence regime. At a higher fluence of ˜10 × 10 23 m -2, on the other hand, an anomalous increase of the coefficient was detected irrespective of the orientation. The results were interpreted as due to competing irradiation mechanisms of the formation of Cu-rich precipitates, recovery process, and the formation of late-blooming Mn-Ni-Si-rich clusters.

  2. Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)

    SciTech Connect

    Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K.

    1992-03-01

    The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the ``primary acceptance criterion`` in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.

  3. Review of reactor pressure vessel evaluation report for Yankee Rowe Nuclear Power Station (YAEC No. 1735)

    SciTech Connect

    Cheverton, R.D.; Dickson, T.L.; Merkle, J.G.; Nanstad, R.K. )

    1992-03-01

    The Yankee Atomic Electric Company has performed an Integrated Pressurized Thermal Shock (IPTS)-type evaluation of the Yankee Rowe reactor pressure vessel in accordance with the PTS Rule (10 CFR 50. 61) and a US Regulatory Guide 1.154. The Oak Ridge National Laboratory (ORNL) reviewed the YAEC document and performed an independent probabilistic fracture-mechnics analysis. The review included a comparison of the Pacific Northwest Laboratory (PNL) and the ORNL probabilistic fracture-mechanics codes (VISA-II and OCA-P, respectively). The review identified minor errors and one significant difference in philosophy. Also, the two codes have a few dissimilar peripheral features. Aside from these differences, VISA-II and OCA-P are very similar and with errors corrected and when adjusted for the difference in the treatment of fracture toughness distribution through the wall, yield essentially the same value of the conditional probability of failure. The ORNL independent evaluation indicated RT{sub NDT} values considerably greater than those corresponding to the PTS-Rule screening criteria and a frequency of failure substantially greater than that corresponding to the primary acceptance criterion'' in US Regulatory Guide 1.154. Time constraints, however, prevented as rigorous a treatment as the situation deserves. Thus, these results are very preliminary.

  4. High pressure storage vessel

    SciTech Connect

    Liu, Qiang

    2013-08-27

    Disclosed herein is a composite pressure vessel with a liner having a polar boss and a blind boss a shell is formed around the liner via one or more filament wrappings continuously disposed around at least a substantial portion of the liner assembly combined the liner and filament wrapping have a support profile. To reduce susceptible to rupture a locally disposed filament fiber is added.

  5. Hybrid Inflatable Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Raboin, Jasen; Valle, Gerard D.; Edeen, Gregg; DeLaFuente, Horacio M.; Schneider, William C.; Spexarth, Gary R.; Johnson, Christopher J.; Pandya, Shalini

    2004-01-01

    Figure 1 shows a prototype of a large pressure vessel under development for eventual use as a habitable module for long spaceflight (e.g., for transporting humans to Mars). The vessel is a hybrid that comprises an inflatable shell attached to a rigid central structural core. The inflatable shell is, itself, a hybrid that comprises (1) a pressure bladder restrained against expansion by (2) a web of straps made from high-strength polymeric fabrics. On Earth, pressure vessels like this could be used, for example, as portable habitats that could be set up quickly in remote locations, portable hyperbaric chambers for treatment of decompression sickness, or flotation devices for offshore platforms. In addition, some aspects of the design of the fabric straps could be adapted to such other items as lifting straps, parachute straps, and automotive safety belts. Figure 2 depicts selected aspects of the design of a vessel of this type with a toroidal configuration. The bladder serves as an impermeable layer to keep air within the pressure vessel and, for this purpose, is sealed to the central structural core. The web includes longitudinal and circumferential straps. To help maintain the proper shape upon inflation after storage, longitudinal and circumferential straps are indexed together at several of their intersections. Because the web is not required to provide a pressure seal and the bladder is not required to sustain structural loads, the bladder and the web can be optimized for their respective functions. Thus, the bladder can be sealed directly to the rigid core without having to include the web in the seal substructure, and the web can be designed for strength. The ends of the longitudinal straps are attached to the ends of the rigid structural core by means of clevises. Each clevis pin is surrounded by a roller, around which a longitudinal strap is wrapped to form a lap seam with itself. The roller is of a large diameter chosen to reduce bending of the fibers in

  6. Pressure vessel design manual

    SciTech Connect

    Moss, D.R.

    1987-01-01

    The first section of the book covers types of loadings, failures, and stress theories, and how they apply to pressure vessels. The book delineates the procedures for designing typical components as well as those for designing large openings in cylindrical shells, ring girders, davits, platforms, bins and elevated tanks. The techniques for designing conical transitions, cone-cylinder intersections, intermediate heads, flat heads, and spherically dished covers are also described. The book covers the design of vessel supports subject to wind and seismic loads and one section is devoted to the five major ways of analyzing loads on shells and heads. Each procedure is detailed enough to size all welds, bolts, and plate thicknesses and to determine actual stresses.

  7. Automatic Magnetic Particle Inspection System for the Bracket Welds of Atucha i Nuclear Power Plant Pressure Vessel

    NASA Astrophysics Data System (ADS)

    Katchadjian, P.; Desimone, C.; Garcia, A.; Antonaccio, C.; Schroeter, F.; Mastroleonardo, P.

    2011-06-01

    The present work refers to the welding inspection of the brackets of Atucha I Nuclear Power Plant's Pressure Vessel (RPV) using the wet fluorescent magnetic particles technique (MT). Due to limited access and high radiation levels in the inspection area, it was necessary to automate the testing and use non conventional magnetization techniques. This paper describes the design and implementation of an automated inspection device and the tests carried out on the mock-up to set up the system. Also, magnetization techniques used are described, explaining in detail the non conventional technique of magnetization by current plates and the use of magnetic field concentrators to increase the field values in the area of interest. Finally, the device mounted on the RPV, used to inspect the bracket's weld, and the results achieved from the inspection are shown.

  8. Irradiation-induced changes of the atomic distributions around the interfaces of carbides in a nuclear reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Toyama, T.; Tsuchiya, N.; Nagai, Y.; Almazouzi, A.; Hatakeyama, M.; Hasegawa, M.; Ohkubo, T.; van Walle, E.; Gerard, R.

    2010-10-01

    Irradiation-induced changes of the atomic distributions of solute and impurity elements around carbides in a reactor pressure vessel steel of a Belgium nuclear power reactor were investigated by laser-assisted local electrode-type three-dimensional atom probe, before and after in-service irradiation of 12 years. Before irradiation, nano-scale Fe-Mn-Cr-Mo carbides were found to be intragranular. The atomic distributions of Mn, Cr and Mo inside the carbide indicate that their concentrations around the inner carbide-matrix interface were enhanced, while a clear segregation of P at the interface was observed. After irradiation, the Mn concentration in the carbide increased substantially. In addition, the enhancement of Mn, Cr and Mo concentrations around the interface and the segregation of P were markedly intensified.

  9. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, James K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

  10. Apollo experience report: Pressure vessels

    NASA Technical Reports Server (NTRS)

    Ecord, G. M.

    1972-01-01

    The Apollo spacecraft pressure vessels, associated problems and resolutions, and related experience in evaluating potential problem areas are discussed. Information is provided that can be used as a guideline in the establishment of baseline criteria for the design and use of lightweight pressure vessels. One of the first practical applications of the use of fracture-mechanics technology to protect against service failures was made on Apollo pressure vessels. Recommendations are made, based on Apollo experience, that are designed to reduce the incidence of failure in pressure-vessel operation and service.

  11. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Sun, Mingyue; Hao, Luhan; Li, Shijian; Li, Dianzhong; Li, Yiyi

    2011-11-01

    Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  12. Graphite filament wound pressure vessels

    NASA Technical Reports Server (NTRS)

    Feldman, A.; Damico, J. J.

    1972-01-01

    Filament wound NOL rings, 4-inch and 8-inch diameter closed-end vessels involving three epoxy resin systems and three graphite fibers were tested to develop property data and fabrication technology for filament wound graphite/epoxy pressure vessels. Vessels were subjected to single-cycle burst tests at room temperature. Manufacturing parameters were established for tooling, winding, and curing that resulted in the development of a pressure/vessel performance factor (pressure x volume/weight) or more than 900,000 in. for an oblate spheroid specimen.

  13. Effects of thermal aging on microstructure and hardness of stainless steel weld-overlay claddings of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.; Suzuki, M.

    2014-09-01

    The effects of thermal aging of stainless steel weld-overlay claddings of nuclear reactor pressure vessels on the microstructure and hardness of the claddings were investigated using atom probe tomography and nanoindentation testing. The claddings were aged at 400 °C for periods of 100-10,000 h. The fluctuation in Cr concentration in the δ-ferrite phase, which was caused by spinodal decomposition, progressed rapidly after aging for 100 h, and gradually for aging durations greater than 1000 h. On the other hand, NiSiMn clusters, initially formed after aging for less than 1000 h, had the highest number density after aging for 2000 h, and coarsened after aging for 10,000 h. The hardness of the δ-ferrite phase also increased rapidly for short period of aging, and saturated after aging for longer than 1000 h. This trend was similar to the observed Cr fluctuation concentration, but different from the trend seen in the formation of the NiSiMn clusters. These results strongly suggest that the primary factor responsible for the hardening of the δ-ferrite phase owing to thermal aging is Cr spinodal decomposition.

  14. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application

    SciTech Connect

    Ren, Weiju; Terry, Totemeier

    2006-10-01

    Two different topics of Grade 91 steel are investigated for Gen IV nuclear reactor pressure vessel application. On the first topic, negligible creep of Grade 91 is investigated with the motivation to design the reactor pressure vessel in negligible creep regime and eliminate costly surveillance programs during the reactor operation. Available negligible creep criteria and creep strain laws are reviewed, and new data needs are evaluated. It is concluded that modifications of the existing criteria and laws, together with their associated parameters, are needed before they can be reliably applied to Grade 91 for negligible creep prediction and reactor pressure vessel design. On the second topic, effects of off-normal welding and heat treatment on creep behavior of Grade 91 are studied with the motivation to better define the control over the parameters in welding and heat treatment procedures. The study is focused on off-normal austenitizing temperatures and improper cooling after welding but prior to post-weld heat treatment.

  15. Multilayer Composite Pressure Vessels

    NASA Technical Reports Server (NTRS)

    DeLay, Tom

    2005-01-01

    A method has been devised to enable the fabrication of lightweight pressure vessels from multilayer composite materials. This method is related to, but not the same as, the method described in gMaking a Metal- Lined Composite-Overwrapped Pressure Vessel h (MFS-31814), NASA Tech Briefs, Vol. 29, No. 3 (March 2005), page 59. The method is flexible in that it poses no major impediment to changes in tank design and is applicable to a wide range of tank sizes. The figure depicts a finished tank fabricated by this method, showing layers added at various stages of the fabrication process. In the first step of the process, a mandrel that defines the size and shape of the interior of the tank is machined from a polyurethane foam or other suitable lightweight tooling material. The mandrel is outfitted with metallic end fittings on a shaft. Each end fitting includes an outer flange that has a small step to accommodate a thin layer of graphite/epoxy or other suitable composite material. The outer surface of the mandrel (but not the fittings) is covered with a suitable release material. The composite material is filament- wound so as to cover the entire surface of the mandrel from the step on one end fitting to the step on the other end fitting. The composite material is then cured in place. The entire workpiece is cut in half in a plane perpendicular to the axis of symmetry at its mid-length point, yielding two composite-material half shells, each containing half of the foam mandrel. The halves of the mandrel are removed from within the composite shells, then the shells are reassembled and bonded together with a belly band of cured composite material. The resulting composite shell becomes a mandrel for the subsequent steps of the fabrication process and remains inside the final tank. The outer surface of the composite shell is covered with a layer of material designed to be impermeable by the pressurized fluid to be contained in the tank. A second step on the outer flange of

  16. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, J.K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell. 6 figures.

  17. Reactor pressure vessel. Status report

    SciTech Connect

    Elliot, B.J.; Hackett, E.M.; Lee, A.D.

    1996-10-01

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff`s reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date.

  18. Carbon fiber internal pressure vessels

    NASA Technical Reports Server (NTRS)

    Simon, R. A.

    1973-01-01

    Internal pressure vessels were designed; the filament was wound of carbon fibers and epoxy resin and tested to burst. The fibers used were Thornel 400, Thornel 75, and Hercules HTS. Additional vessels with type A fiber were made. Polymeric linears were used, and all burst testing was done at room temperature. The objective was to produce vessels with the highest attainable PbV/W efficiencies. The type A vessels showed the highest average efficiency: 2.56 x 10 to the 6th power cm. Next highest efficiency was with Thornel 400 vessels: 2.21 x 10 to the 6th power cm. These values compare favorably with efficiency values from good quality S-glass vessels, but strains averaged 0.97% or less, which is less than 1/3 the strain of S-glass vessels.

  19. Special enclosure for a pressure vessel

    SciTech Connect

    Wedellsborg, B.W.; Wedellsborg, U.W.

    1993-06-08

    A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom.

  20. Level indicator for pressure vessels

    DOEpatents

    Not Available

    1982-04-28

    A liquid-level monitor for tracking the level of a coal slurry in a high-pressure vessel including a toroidal-shaped float with magnetically permeable bands thereon disposed within the vessel, two pairs of magnetic-field generators and detectors disposed outside the vessel adjacent the top and bottom thereof and magnetically coupled to the magnetically permeable bands on the float, and signal-processing circuitry for combining signals from the top and bottom detectors for generating a monotonically increasing analog control signal which is a function of liquid level. The control signal may be utilized to operate high-pressure control valves associated with processes in which the high-pressure vessel is used.

  1. Cuff for Blood-Vessel Pressure Measurements

    NASA Technical Reports Server (NTRS)

    Shimizu, M.

    1982-01-01

    Pressure within blood vessel is measured by new cufflike device without penetration of vessel. Device continuously monitors blood pressure for up to 6 months or longer without harming vessel. Is especially useful for vessels smaller than 4 or 5 millimeters in diameter. Invasive methods damage vessel wall, disturb blood flow, and cause clotting. They do not always give reliable pressure measurements over prolonged periods.

  2. Cuff for Blood-Vessel Pressure Measurements

    NASA Technical Reports Server (NTRS)

    Shimizu, M.

    1982-01-01

    Pressure within blood vessel is measured by new cufflike device without penetration of vessel. Device continuously monitors blood pressure for up to 6 months or longer without harming vessel. Is especially useful for vessels smaller than 4 or 5 millimeters in diameter. Invasive methods damage vessel wall, disturb blood flow, and cause clotting. They do not always give reliable pressure measurements over prolonged periods.

  3. 46 CFR 169.249 - Pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 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...

  4. 46 CFR 182.330 - Pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 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)...

  5. 46 CFR 169.249 - Pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 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.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 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. 46 CFR 169.249 - Pressure vessels.

    Code of Federal Regulations, 2010 CFR

    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...

  8. 46 CFR 182.330 - Pressure vessels.

    Code of Federal Regulations, 2010 CFR

    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)...

  9. 46 CFR 182.330 - Pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 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)...

  10. 46 CFR 169.249 - Pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 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...

  11. 46 CFR 182.330 - Pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 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)...

  12. 46 CFR 169.249 - Pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 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...

  13. Pressure vessel having continuous sidewall

    NASA Technical Reports Server (NTRS)

    Simon, Xavier D. (Inventor); Barackman, Victor J. (Inventor)

    2011-01-01

    A spacecraft pressure vessel has a tub member. A sidewall member is coupled to the tub member so that a bottom section of the sidewall member extends from an attachment intersection with the tub member and away from the tub member. The bottom section of the sidewall member receives and transfers a load through the sidewall member.

  14. (Irradiation embrittlement of reactor pressure vessels)

    SciTech Connect

    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.

  15. 46 CFR 119.330 - Pressure vessels.

    Code of Federal Regulations, 2010 CFR

    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...

  16. 46 CFR 119.330 - Pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 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...

  17. 46 CFR 119.330 - Pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 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...

  18. 46 CFR 119.330 - Pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 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...

  19. 46 CFR 119.330 - Pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 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...

  20. Three-term Asymptotic Stress Field Expansion for Analysis of Surface Cracked Elbows in Nuclear Pressure Vessels

    NASA Astrophysics Data System (ADS)

    Labbe, Fernando

    2007-04-01

    Elbows with a shallow surface cracks in nuclear pressure pipes have been recognized as a major origin of potential catastrophic failures. Crack assessment is normally performed by using the J-integral approach. Although this one-parameter-based approach is useful to predict the ductile crack onset, it depends strongly on specimen geometry or constraint level. When a shallow crack exists (depth crack-to-thickness wall ratio less than 0.2) and/or a fully plastic condition develops around the crack, the J-integral alone does not describe completely the crack-tip stress field. In this paper, we report on the use of a three-term asymptotic expansion, referred to as the J- A 2 methodology, for modeling the elastic-plastic stress field around a three-dimensional shallow surface crack in an elbow subjected to internal pressure and out-of-plane bending. The material, an A 516 Gr. 70 steel, used in the nuclear industry, was modeled with a Ramberg-Osgood power law and flow theory of plasticity. A finite deformation theory was included to account for the highly nonlinear behavior around the crack tip. Numerical finite element results were used to calculate a second fracture parameter A 2 for the J- A 2 methodology. We found that the used three-term asymptotic expansion accurately describes the stress field around the considered three-dimensional shallow surface crack.

  1. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Code of Federal Regulations, 2010 CFR

    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...

  2. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 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...

  3. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 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...

  4. 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

    SciTech Connect

    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.

  5. Steel pressure vessels for hydrostatic pressures to 50 kilobars.

    PubMed

    Lavergne, A; Whalley, E

    1978-07-01

    Cylindrical steel pressure vessels are described that can be used for hydrostatic pressures up to 50 kilobars. Monoblock vessels of 350 maraging steel can be used to 40 kilobars and compound vessels with an inner vessel of 350 maraging steel and an outer vessel of 300 maraging steel to 50 kilobars. Neither requires the cylinder to be end loaded, and so they are much easier to use than the more usual compound vessels with a tungsten carbide inner and steel outer vessel.

  6. Reactor Pressure Vessel (RPV) Acquisition Strategy

    SciTech Connect

    Mizia, Ronald Eugene

    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

  7. Microstructural changes of a thermally aged stainless steel submerged arc weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kameda, J.; Nagai, Y.; Toyama, T.; Matsukawa, Y.; Nishiyama, Y.; Onizawa, K.

    2012-06-01

    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 δ-ferrite phase but not in the austenitic phase. Thermal aging at 400 °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 δ-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 γ-austenite and δ-ferrite interface. There were no Cr depleted zones around the carbide.

  8. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K.

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 1019 n cm-2 (E > 1 MeV) and a flux of 1.1 × 1013 n cm-2 s-1 at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  9. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    NASA Astrophysics Data System (ADS)

    Takeuchi, T.; Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y.; Nishiyama, Y.; Katsuyama, J.; Onizawa, K.; Suzuki, M.

    2013-11-01

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 1023 n m-2 (E > 1 MeV) and a flux of 1.1 × 1017 n m-2 s-1 at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ‧-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  10. Filament wound pressure vessels - Effects of using liner tooling of low pressure vessels for high pressure vessels development

    NASA Astrophysics Data System (ADS)

    Lal, Krishna M.

    High performance pressure vessels have been recently demanded for aerospace and defense applications. Filament wound pressure vessels consist of a metallic thin liner, which also acts as a mandrel, and composite/epoxy overwrap. Graphite/epoxy overwrapped vessels have been developed to obtain the performance ratio, PV/W, as high as one million inches. Under very high pressure the isotropic metallic liner deforms elasto-plastically, and orthotropic composite fibers deform elastically. Sometimes, for the development of ultra high pressure vessels, composite pressure vessels industry uses the existing liner tooling developed for low burst pressure capacity composite vessels. This work presents the effects of various design variables including the low pressure liner tooling for the development of the high burst pressure capacity Brilliant Pebbles helium tanks. Advance stress analysis and development of an ultra high pressure helium tank.

  11. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    SciTech Connect

    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.

  12. THERMAL ANNEALING OF REACTOR PRESSURE VESSELS

    SciTech Connect

    Sokolov, Mikhail A; Server, W. L.; Nanstad, Randy K

    2015-01-01

    Some of the current fleet of nuclear power plants is poised to reach their end of life and will require an operating life time extension. Therefore, the main structural components, including the reactor pressure vessel (RPV), will be subject to higher neutron exposures than originally planned. These longer operating times raise serious concerns regarding our ability to manage the reliability of RPV steels at such high doses. Thermal annealing is the only option that can, to some degree, recover irradiated beltline region transition temperature shift and recover upper shelf energy properties lost during radiation exposure and extend RPV service life. This paper reviews the experience accumulated internationally with development and implementation of thermal annealing to RPV and potential perspectives for carrying out thermal annealing on US nuclear power plant RPVs.

  13. Hydrogen storage in insulated pressure vessels

    SciTech Connect

    Aceves, S.M.; Garcia-Villazana, O.

    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.

  14. LPT. EBOR reactor vessel in TAN 646. Pressure vessel head ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    LPT. EBOR reactor vessel in TAN 646. Pressure vessel head being installed in vault. Refueling port extension (right) and control rod nozzles (center). Camera facing northwest. Photographer: Comiskey. Date: January 20, 1965. INEEL negative no. 65-241 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  15. 46 CFR 4.03-35 - Nuclear vessel.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... INVESTIGATIONS Definitions § 4.03-35 Nuclear vessel. The term nuclear vessel means any vessel in which power for propulsion, or for any other purpose, is derived from nuclear energy; or any vessel handling or processing... 46 Shipping 1 2014-10-01 2014-10-01 false Nuclear vessel. 4.03-35 Section 4.03-35 Shipping...

  16. 46 CFR 4.03-35 - Nuclear vessel.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... INVESTIGATIONS Definitions § 4.03-35 Nuclear vessel. The term nuclear vessel means any vessel in which power for propulsion, or for any other purpose, is derived from nuclear energy; or any vessel handling or processing... 46 Shipping 1 2011-10-01 2011-10-01 false Nuclear vessel. 4.03-35 Section 4.03-35 Shipping...

  17. 46 CFR 4.03-35 - Nuclear vessel.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... INVESTIGATIONS Definitions § 4.03-35 Nuclear vessel. The term nuclear vessel means any vessel in which power for propulsion, or for any other purpose, is derived from nuclear energy; or any vessel handling or processing... 46 Shipping 1 2012-10-01 2012-10-01 false Nuclear vessel. 4.03-35 Section 4.03-35 Shipping...

  18. 46 CFR 4.03-35 - Nuclear vessel.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... INVESTIGATIONS Definitions § 4.03-35 Nuclear vessel. The term nuclear vessel means any vessel in which power for propulsion, or for any other purpose, is derived from nuclear energy; or any vessel handling or processing... 46 Shipping 1 2010-10-01 2010-10-01 false Nuclear vessel. 4.03-35 Section 4.03-35 Shipping...

  19. 46 CFR 4.03-35 - Nuclear vessel.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... INVESTIGATIONS Definitions § 4.03-35 Nuclear vessel. The term nuclear vessel means any vessel in which power for propulsion, or for any other purpose, is derived from nuclear energy; or any vessel handling or processing... 46 Shipping 1 2013-10-01 2013-10-01 false Nuclear vessel. 4.03-35 Section 4.03-35 Shipping...

  20. Wrapped Wire Detects Rupture Of Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Hunt, James B.

    1990-01-01

    Simple, inexpensive technique helps protect against damage caused by continuing operation of equipment after rupture or burnout of pressure vessel. Wire wrapped over area on outside of vessel where breakthrough most likely. If wall breaks or burns, so does wire. Current passing through wire ceases, triggering cutoff mechanism stopping flow in vessel to prevent further damage. Applied in other situations in which pipes or vessels fail due to overpressure, overheating, or corrosion.

  1. Wrapped Wire Detects Rupture Of Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Hunt, James B.

    1990-01-01

    Simple, inexpensive technique helps protect against damage caused by continuing operation of equipment after rupture or burnout of pressure vessel. Wire wrapped over area on outside of vessel where breakthrough most likely. If wall breaks or burns, so does wire. Current passing through wire ceases, triggering cutoff mechanism stopping flow in vessel to prevent further damage. Applied in other situations in which pipes or vessels fail due to overpressure, overheating, or corrosion.

  2. Lightweight bladder lined pressure vessels

    DOEpatents

    Mitlitsky, Fred; Myers, Blake; Magnotta, Frank

    1998-01-01

    A lightweight, low permeability liner for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using torispherical or near torispherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film seamed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life.

  3. Lightweight bladder lined pressure vessels

    DOEpatents

    Mitlitsky, F.; Myers, B.; Magnotta, F.

    1998-08-25

    A lightweight, low permeability liner is described for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using tori spherical or near tori spherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film sealed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life. 19 figs.

  4. Reactor pressure vessel structural integrity research

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    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.

  5. Pressure vessel burst test program. II

    NASA Technical Reports Server (NTRS)

    Cain, Maurice R.; Sharp, Douglas E.; Coleman, Michael D.

    1991-01-01

    The current status is disucssed of a program to study the characteristics of blast waves and fragmentation generated by ruptured gas-filled pressure vessels. Current methods for assessing vessel safety and burst parameters are briefly reviewed, and pneumatic burst testing operations and testing results are examined. A comparison is made with current methods for burst assessment. It is tentatively concluded that, at close distances, vessel burst overpressures are less than those of high-explosive (HE) blasts with equivalent energy and are greater than HE far from the vessel. The impulse appears to be the same for both vessel bursts and equivalent energy HE blasts. The functional relationship between shock velocity and overpressure ratio appears to be the same for vessel bursts as for HE blasts. The initial shock overpressure appears to be much less than vessel pressure and may be found using the one-dimensional shock tube equation.

  6. Detection of irradiation embrittlement of low-alloy steel for nuclear reactor pressure vessels using a probe type eddy current sensor

    SciTech Connect

    Maeda, Noriyoshi; Yamaguchi, Atsunori; Sugibayashi, Takuya; Kohno, Katsumi

    1999-10-01

    This report describes the results of studies made for the purpose of detecting the irradiation embrittlement of low-alloy steel used for nuclear reactor pressure vessels. For the method of using eddy current to detect material degradation, the device and the sensor employed are light in weight and compact in size, allowing testing without contact. In this study the frequency of input current to the excitation coil is changed in steps of 1 kHz. The output signal is processed by phase detection method, and displayed on a complex plane. It depicts a trajectory as the frequency is changed. To extract features of the trajectories, averaged radius and averaged phase angle are defined and plotted as function of neutron fluence or ductile-brittle transition temperature. Experiment shows that the averaged phase angle and transition temperature decrease as the neutron fluence is increased. Behavior of the averaged phase angle is interpreted employing magnetic permeability and electric conductivity of the test specimens. It becomes clear that electric conductivity decreases as the neutron fluence is increased.

  7. 46 CFR 4.05-35 - Incidents involving nuclear vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 1 2012-10-01 2012-10-01 false Incidents involving nuclear vessels. 4.05-35 Section 4... involving nuclear vessels. The master of any nuclear vessel shall immediately inform the Commandant in the event of any accident or casualty to the nuclear vessel which may lead to an environmental hazard. The...

  8. 46 CFR 4.05-35 - Incidents involving nuclear vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 1 2014-10-01 2014-10-01 false Incidents involving nuclear vessels. 4.05-35 Section 4... involving nuclear vessels. The master of any nuclear vessel shall immediately inform the Commandant in the event of any accident or casualty to the nuclear vessel which may lead to an environmental hazard. The...

  9. 46 CFR 4.05-35 - Incidents involving nuclear vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 1 2013-10-01 2013-10-01 false Incidents involving nuclear vessels. 4.05-35 Section 4... involving nuclear vessels. The master of any nuclear vessel shall immediately inform the Commandant in the event of any accident or casualty to the nuclear vessel which may lead to an environmental hazard. The...

  10. 46 CFR 4.05-35 - Incidents involving nuclear vessels.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Incidents involving nuclear vessels. 4.05-35 Section 4... involving nuclear vessels. The master of any nuclear vessel shall immediately inform the Commandant in the event of any accident or casualty to the nuclear vessel which may lead to an environmental hazard....

  11. 46 CFR 4.05-35 - Incidents involving nuclear vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... involving nuclear vessels. The master of any nuclear vessel shall immediately inform the Commandant in the event of any accident or casualty to the nuclear vessel which may lead to an environmental hazard. The... 46 Shipping 1 2011-10-01 2011-10-01 false Incidents involving nuclear vessels. 4.05-35 Section 4...

  12. Method of manufacturing an overwrapped pressure vessel

    NASA Technical Reports Server (NTRS)

    Beck, Emory J. (Inventor)

    1976-01-01

    A pressure vessel of the type wherein a metallic liner in the shape of a cylindrical portion with a dome-shaped portion at each end thereof is overwrapped by a plurality of layers of resin coated, single fiberglass filaments. A four-step wrapping technique reinforces the vessel with overwrap material at the most likely areas for vessel failure. Overwrapping of the vessel is followed by a sizing pressurization cycle which induces a compressive prestress into the liner and thereby permits the liner to deform elastically through an increased strain range.

  13. Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

    SciTech Connect

    Wang, J.A.; Kam, F.B.K.

    1998-02-01

    The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs.

  14. HTGR Base Technology Program. Task 2: concrete properties in nuclear environment. A review of concrete material systems for application to prestressed concrete pressure vessels

    SciTech Connect

    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.

  15. Nuclear reactor vessel fuel thermal insulating barrier

    DOEpatents

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  16. Feedthrough Seal For High-Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Williams, R.; Mullins, O.; Smith, D.; Teasley, G.

    1984-01-01

    Combination of ceramic and plastic withstands many depressurizations. Stack of washers surrounds leadthrough electrode. Under pressure washers expand to fill leadthrough hole in high-pressure vessel. Seal thus formed withstands 20 or more pressurization/depressurization cycles. Seal composed of neoprene, polytetrafluoroethylene, nylon and high-purity, high-density commercial alumina ceramic.

  17. Steel - Structural, reinforcing; Pressure vessel, railway

    SciTech Connect

    Not Available

    1986-01-01

    This book contains specifications for structural steel used in various constructions; concrete reinforcement; plate and forgings for boilers and pressure vesseles; rails, axles, wheels and other accessories for railway service.

  18. Radiation effects on reactor pressure vessel supports

    SciTech Connect

    Johnson, R.E.; Lipinski, R.E.

    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.

  19. Discontinuity stresses in metallic pressure vessels

    NASA Technical Reports Server (NTRS)

    1971-01-01

    The state of the art, criteria, and recommended practices for the theoretical and experimental analyses of discontinuity stresses and their distribution in metallic pressure vessels for space vehicles are outlined. The applicable types of pressure vessels include propellant tanks ranging from main load-carrying integral tank structure to small auxiliary tanks, storage tanks, solid propellant motor cases, high pressure gas bottles, and pressurized cabins. The major sources of discontinuity stresses are discussed, including deviations in geometry, material properties, loads, and temperature. The advantages, limitations, and disadvantages of various theoretical and experimental discontinuity analysis methods are summarized. Guides are presented for evaluating discontinuity stresses so that pressure vessel performance will not fall below acceptable levels.

  20. Neutron shielding panels for reactor pressure vessels

    SciTech Connect

    Singleton, Norman R

    2011-11-22

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  1. Flexible Composite-Material Pressure Vessel

    NASA Technical Reports Server (NTRS)

    Brown, Glen; Haggard, Roy; Harris, Paul A.

    2003-01-01

    A proposed lightweight pressure vessel would be made of a composite of high-tenacity continuous fibers and a flexible matrix material. The flexibility of this pressure vessel would render it (1) compactly stowable for transport and (2) more able to withstand impacts, relative to lightweight pressure vessels made of rigid composite materials. The vessel would be designed as a structural shell wherein the fibers would be predominantly bias-oriented, the orientations being optimized to make the fibers bear the tensile loads in the structure. Such efficient use of tension-bearing fibers would minimize or eliminate the need for stitching and fill (weft) fibers for strength. The vessel could be fabricated by techniques adapted from filament winding of prior composite-material vessels, perhaps in conjunction with the use of dry film adhesives. In addition to the high-bias main-body substructure described above, the vessel would include a low-bias end substructure to complete coverage and react peak loads. Axial elements would be overlaid to contain damage and to control fiber orientation around side openings. Fiber ring structures would be used as interfaces for connection to ancillary hardware.

  2. In-service pressure vessel inspections

    NASA Astrophysics Data System (ADS)

    Fields, Marvin

    1998-03-01

    Eliminate any doubt that the vessel condition is suitable for continued operation through a planned inspection program that can mitigate or avoid failure of a pressure vessel due to corrosion or erosion. Proper inspection and documentation help you in identifying the problem and confirming the actual thickness leading to properly correcting deficiencies. Proper inspection is the antidote for any inspection program. Vessel life can be extended, risk can be minimized and unscheduled downtime can be prevented by implementing and managing your inspection program. A successful program includes maintaining accurate records, conducting inspections in regular intervals, and taking proper action on deficiencies. Therefore, you will know what you have and the condition of your equipment. Pressure vessel inspections can be classified into two general categories: surface inspection and volumetric inspection. Surface techniques for vessels include two of the commonest types: dye-penetrant and magnetic particle testing. Board qualified inspectors are required to perform these two tests. Volumetric techniques for vessels include three common types: ultrasonic testing, eddy current testing, and radiography. At Abbott the use of advanced NDE (non destructive examination) techniques, ultrasonic b-scan, has provided us with the proper tools to obtain the above objectives. We have been applying ultrasonic b-scan utilizing a pulse echo pitch catch technique to provide us with essential data on each of our pressure vessels. This reduces equipment downtime because the nondestructive examination usually takes place while our vessels are in service. As inspections take place we are able to view a real time image of the defective discontinuities on a video monitor. This ultrasonic b-scan technique is allowing us to perform fast accurate examinations covering up to 96% of the surface area of each pressure vessel.

  3. Liquid Nitrogen Subcooler Pressure Vessel Engineering Note

    SciTech Connect

    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.

  4. Nickel hydrogen common pressure vessel battery development

    NASA Technical Reports Server (NTRS)

    Jones, Kenneth R.; Zagrodnik, Jeffrey P.

    1992-01-01

    Our present design for a common pressure vessel (CPV) battery, a nickel hydrogen battery system to combine all of the cells into a common pressure vessel, uses an open disk which allows the cell to be set into a shallow cavity; subsequent cells are stacked on each other with the total number based on the battery voltage required. This approach not only eliminates the assembly error threat, but also more readily assures equal contact pressure to the heat fin between each cell, which further assures balanced heat transfer. These heat fin dishes with their appropriate cell stacks are held together with tie bars which in turn are connected to the pressure vessel weld rings at each end of the tube.

  5. Progress Report: Pressure Vessel Burst Test Study

    DTIC Science & Technology

    1994-08-01

    report is provided on a program developed to study through test and analysis, the characteristics of blast waves and fragmentation generated by ruptured ...vessels were composite overwrapped pressure vessels ( COPV ) and were cut with a shaped charge (no groove) around its center. The burst location on the...and the shaped charge cut area (shown with dotted lines). BURST INITIATION Longitudinal stress in the circumferential grooves (for developing axial

  6. Design Considerations For Blast Loads In Pressure Vessels.

    SciTech Connect

    Rodriguez, E. A.; Nickell, Robert E.; Pepin, J. 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.

  7. Blood vessels, circulation and blood pressure.

    PubMed

    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.

  8. Mechanical characteristics of filament-wound pressure vessel (burst pressure)

    NASA Technical Reports Server (NTRS)

    Iida, H.; Uemura, M.

    1987-01-01

    The finite element method is used to analyze the mechanical characteristics of a pressurized filament-wound (FW) pressure vessel, and to predict its burst pressure. The analysis takes into account the bending moment, the stretch-bend coupling effect, nonlinear stress-strain relations, and finite deflection. The analysis is based on two initial failure criteria for laminae, and two ultimate fracture criteria for laminated structures. The numerical results, obtained by applying the load incremental method to the isotensoid CFRP pressure vessel used in the launching of the Zikiken satellite, are in good agreement with the experimental burst pressure and fracture behaviors.

  9. Mechanical characteristics of filament-wound pressure vessel (burst pressure)

    NASA Technical Reports Server (NTRS)

    Iida, H.; Uemura, M.

    1987-01-01

    The finite element method is used to analyze the mechanical characteristics of a pressurized filament-wound (FW) pressure vessel, and to predict its burst pressure. The analysis takes into account the bending moment, the stretch-bend coupling effect, nonlinear stress-strain relations, and finite deflection. The analysis is based on two initial failure criteria for laminae, and two ultimate fracture criteria for laminated structures. The numerical results, obtained by applying the load incremental method to the isotensoid CFRP pressure vessel used in the launching of the Zikiken satellite, are in good agreement with the experimental burst pressure and fracture behaviors.

  10. 46 CFR 61.10-5 - Pressure vessels in service.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 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...

  11. 46 CFR 61.10-5 - Pressure vessels in service.

    Code of Federal Regulations, 2010 CFR

    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...

  12. 46 CFR 61.10-5 - Pressure vessels in service.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 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...

  13. 46 CFR 61.10-5 - Pressure vessels in service.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 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...

  14. 46 CFR 50.30-15 - Class II pressure vessels.

    Code of Federal Regulations, 2010 CFR

    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...

  15. 46 CFR 61.10-5 - Pressure vessels in service.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 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...

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

    Code of Federal Regulations, 2010 CFR

    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. Curved and conformal high-pressure vessel

    SciTech Connect

    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.

  18. Thermomechanical fracture on pressurized cylindrical vessels

    NASA Astrophysics Data System (ADS)

    Tzou, Robert D. Y.; Chiu, Kwong S.; Beraun, Jorge E.; Chen, Jinn Kuen

    1998-09-01

    This work studies the rapid fracture developed on the surface of a pressurized cylindrical vessel when heated by an intensified energy source. The primary concerns are the interactions between the rapid thermal expansion and the internal pressure that exerts on the interior surface. From a mechanical point of view, the thermal loading tends to develop a crack along the circumferential direction of the cylindrical vessel. The excessive internal pressure established within the cylindrical vessel, on the other hand, tends to develop a crack in the axial direction. Combination of the two mechanisms results in a capricious pattern of rapid fracture that needs to be fully understood in thermal processing. Special features in this work include the dynamics plasticity induced by the combined thermomechanical loading at short times, as well as the temperature-dependent thermomechanical properties that evolve in the load-time history.

  19. Proactive life extension of pressure vessels

    NASA Astrophysics Data System (ADS)

    Mager, Lloyd

    1998-03-01

    For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

  20. Guidelines for pressure vessel safety assessment

    NASA Astrophysics Data System (ADS)

    Yukawa, S.

    1990-04-01

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

  1. Cavity closure arrangement for high pressure vessels

    DOEpatents

    Amtmann, Hans H.

    1981-01-01

    A closure arrangement for a pressure vessel such as the pressure vessel of a high temperature gas-cooled reactor wherein a liner is disposed within a cavity penetration in the reactor vessel and defines an access opening therein. A closure is adapted for sealing relation with an annular mounting flange formed on the penetration liner and has a plurality of radially movable locking blocks thereon having outer serrations adapted for releasable interlocking engagement with serrations formed internally of the upper end of the penetration liner so as to effect high strength closure hold-down. In one embodiment, ramping surfaces are formed on the locking block serrations to bias the closure into sealed relation with the mounting flange when the locking blocks are actuated to locking positions.

  2. Reactor pressure vessel with forged nozzles

    DOEpatents

    Desai, Dilip R.

    1993-01-01

    Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

  3. 46 CFR 115.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 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...

  4. 46 CFR 115.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 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...

  5. 46 CFR 115.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 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...

  6. 46 CFR 115.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 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. High-pressure cryogenic seals for pressure vessels

    NASA Technical Reports Server (NTRS)

    Buggele, A. E.

    1977-01-01

    This investigation of the problems associated with reliably containing gaseous helium pressurized to 1530 bars (22 500 psi) between 4.2 K and 150 K led to the following conclusions: (1) common seal designs used in existing elevated-temperature pressure vessels are unsuitable for high-pressure cryogenic operation, (2) extrusion seal-ring materials such as Teflon, tin, and lead are not good seal materials for cryogenic high-pressure operation; and (3) several high-pressure cryogenic seal systems suitable for large-pressure vessel applications were developed; two seals required prepressurization, and one seal functioned repeatedly without any prepressurization. These designs used indium seal rings, brass or 304 stainless-steel anvil rings, and two O-rings of silicone rubber or Kel-F.

  8. Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study

    SciTech Connect

    Pareige, P.; Russell, K.F.; Stoller, R.E.; Miller, M.K.

    1998-03-01

    Atom probe field ion microscopy (APFIM) investigations of the microstructure of unaged (as-fabricated) and long-term thermally aged ({approximately} 100,000 h at 280 C) surveillance materials from commercial reactor pressure vessel steels were performed. This combination of materials and conditions permitted the investigation of potential thermal-aging effects. This microstructural study focused on the quantification of the compositions of the matrix and carbides. The APFIM results indicate that there was no significant microstructural evolution after a long-term thermal exposure in weld, plate, or forging materials. The matrix depletion of copper that was observed in weld materials was consistent with the copper concentration in the matrix after the stress-relief heat treatment. The compositions of cementite carbides aged for 100,000 h were compared with the Thermocalc{trademark} prediction. The APFIM comparisons of materials under these conditions are consistent with the measured change in mechanical properties such as the Charpy transition temperature.

  9. Development of a methodology for the assessment of shallow-flaw fracture in nuclear reactor pressure vessels: Generation of biaxial shallow-flaw fracture toughness data

    SciTech Connect

    McAfee, W.J.; Bass, B.R.; Bryson, J.W.

    1998-07-01

    A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow-surface flaws. Shallow-flaw fracture toughness of RPV material has been shown to be higher than that for deep flaws, because of the relaxation of crack-tip constraint. This report describes the preliminary test results for a series of cruciform specimens with a uniform depth surface flaw. These specimens are all of the same size with the same depth flaw. Temperature and biaxial load ratio are the independent variables. These tests demonstrated that biaxial loading could have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. Through that temperature range, the effect of full biaxial (1:1) loading on uniaxial, shallow-flaw toughness varied from no effect near the lower shelf to a reduction of approximately 58% at higher temperatures.

  10. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  11. Composite Tanks and Pressure Vessel Development

    NASA Technical Reports Server (NTRS)

    DeLay, Thomas K.; Munafo, Paul M. (Technical Monitor)

    2002-01-01

    Pressure vessels and tanks are vital to NASA missions. Tanks need to be lightweight and perform under the operational environments. Design and material limitations make it difficult to contain the fuels and oxidizers. Recent interest in 90% Hydrogen Peroxide adds to the challenge of containment. The majority of current tank technologies are not easily adaptable to conformal shapes. The cost of tooling-up for large tanks are magnified by sudden design changes. New launch vehicle concepts may require tanks and pressure vessels of a non-standard configuration. Scaled versions of new tanks have been fabricated and testing has begun. Second and third generation launch vehicles decisions will effect the path of research and development.

  12. Guide for certifying pressure vessels and systems

    NASA Technical Reports Server (NTRS)

    Lundy, Floyd; Krusa, Paul W.

    1992-01-01

    This guide is intended to provide methodology and describe the intent of the Pressure Vessel and System (PV/S) Certification program. It is not meant to be a mandated document, but is intended to transmit a basic understanding of the PV/S program, and include examples. After the reader has familiarized himself with this publication, he should have a basic understanding of how to go about developing a PV/S certification program.

  13. Composite Pressure Vessel Including Crack Arresting Barrier

    NASA Technical Reports Server (NTRS)

    DeLay, Thomas K. (Inventor)

    2013-01-01

    A pressure vessel includes a ported fitting having an annular flange formed on an end thereof and a tank that envelopes the annular flange. A crack arresting barrier is bonded to and forming a lining of the tank within the outer surface thereof. The crack arresting barrier includes a cured resin having a post-curing ductility rating of at least approximately 60% through the cured resin, and further includes randomly-oriented fibers positioned in and throughout the cured resin.

  14. Metallized coatings for pressure vessel corrosion

    SciTech Connect

    Hankirk, M. ); Hansen, D.S. )

    1994-09-01

    Metallized coatings have been successful for many years in providing sacrificial protection to pressure vessels in high-temperature applications in which they are susceptible to localized corrosion, hydrogen blistering, erosion, and pitting. In addition, when corrosion allowances have decreased or have been eliminated after many years of service, metallized coatings can be used to restore the allowances and extend the life of the equipment.

  15. Conformable pressure vessel for high pressure gas storage

    DOEpatents

    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.

  16. Compact insert design for cryogenic pressure vessels

    DOEpatents

    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.

  17. Modeling Scala Media as a Pressure Vessel

    NASA Astrophysics Data System (ADS)

    Lepage, Eric; Olofsson, A.˚Ke

    2011-11-01

    The clinical condition known as endolymphatic hydrops is the swelling of scala media and may result in loss in hearing sensitivity consistent with other forms of low-frequency biasing. Because outer hair cells (OHCs) are displacement-sensitive and hearing levels tend to be preserved despite large changes in blood pressure and CSF pressure, it seems unlikely that the OHC respond passively to changes in static pressures in the chambers. This suggests the operation of a major feedback control loop which jointly regulates homeostasis and hearing sensitivity. Therefore the internal forces affecting the cochlear signal processing amplifier cannot be just motile responses. A complete account of the cochlear amplifier must include static pressures. To this end we have added a third, pressure vessel to our 1-D 140-segment, wave-digital filter active model of cochlear mechanics, incorporating the usual nonlinear forward transduction. In each segment the instantaneous pressure is the sum of acoustic pressure and global static pressure. The object of the model is to maintain stable OHC operating point despite any global rise in pressure in the third chamber. Such accumulated pressure is allowed to dissipate exponentially. In this first 3-chamber implementation we explore the possibility that acoustic pressures are rectified. The behavior of the model is critically dependent upon scaling factors and time-constants, yet by initial assumption, the pressure tends to accumulate in proportion to sound level. We further explore setting of the control parameters so that the accumulated pressure either stays within limits or may rise without bound.

  18. Pressure vessel burst test program - Progress paper No. 4

    NASA Technical Reports Server (NTRS)

    Cain, Maurice R.; Sharp, Douglas E.

    1993-01-01

    A status report is presented for a program studying the characteristics of the blast waves and fragmentation caused by ruptured gas-filled pressure vessels. Experimental data trends have been derived from 14 burst pressure vessels. Attention is given to energy release in bursting, blast wave and fragmentation behavior, height of burst effects, fragment velocity vs vessel pressure, and comparative blast effects for spherical/composite vs cylindrical/steel pressure vessels.

  19. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing or to operate the nuclear power reactor following the annealing must be identified. The... licensee shall so confirm in writing to the Director, Office of Nuclear Reactor Regulation. The...

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

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... thermal annealing or to operate the nuclear power reactor following the annealing must be identified. The... licensee shall so confirm in writing to the Director, Office of Nuclear Reactor Regulation. The...

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

    SciTech Connect

    Wang, Jy-An John

    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.

  2. Concept of Small Sized Integrated PWR with Double Pressure Vessels

    SciTech Connect

    Kinoshita, I.; Ueda, N.; Nishi, Y.; Matsumura, T.

    2002-07-01

    For early deployment of small sized nuclear reactors, it is better to reduce the BOP cost with new ideas than introducing innovative technologies for core, fuel and materials. In this report, a concept of the integrated, forced convective and small PWR with double pressure vessels has been proposed. The electric output of this reactor is 150 MW. Conventional technologies are adopted for core and fuel. Refueling, maintenance and repairing are made in a special ship with complete facilities and skilled experts. The pressure vessel with the core, control rod drive mechanisms (CRDM), main circulating pumps (MCP), steam generators (SG) and other reactor internals are transferred between the reactor building and the ship. Technical feasibility for safety and maintainability has been discussed qualitatively. The construction cost has been roughly estimated. (authors)

  3. An evolution of understanding of reactor pressure vessel steel embrittlement

    NASA Astrophysics Data System (ADS)

    Lucas, G. E.

    2010-12-01

    This paper attempts to summarize the lifetime contributions of Prof. G. Robert Odette to our understanding of the effects of neutron irradiation on reactor pressure vessel steel embrittlement. These contributions span the entire range of phenomena that contribute to embrittlement, from the production and evolution of fine scale features by radiation damage processes, to the effects of this damage microstructure on mechanical properties. They include the development and application of unique and novel experimental tools (from Seebeck Coefficient to Small Angle Neutron Scattering to confocal microscopy and fracture reconstruction), the design and implementation of large multi-variable experimental matrices, the application of multiscale modeling to understand the underlying mechanisms of defect evolution and property change, and the development of predictive methodologies employed to govern reactor operations. The ideas and discoveries have provided guidance worldwide to improving the safety of operating nuclear reactor pressure vessels.

  4. 46 CFR 58.60-3 - Pressure vessel.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 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...

  5. 46 CFR 176.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a) Pressure vessels must be tested and inspected in accordance with part 61, subpart 61.10, of this...

  6. 46 CFR 58.60-3 - Pressure vessel.

    Code of Federal Regulations, 2010 CFR

    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...

  7. 46 CFR 58.60-3 - Pressure vessel.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 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...

  8. 46 CFR 176.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a) Pressure vessels must be tested and inspected in accordance with part 61, subpart 61.10, of this...

  9. 46 CFR 58.60-3 - Pressure vessel.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 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...

  10. 46 CFR 176.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2010 CFR

    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) Pressure vessels must be tested and inspected in accordance with part 61, subpart 61.10, of this...

  11. 46 CFR 176.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a) Pressure vessels must be tested and inspected in accordance with part 61, subpart 61.10, of this...

  12. 46 CFR 176.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a) Pressure vessels must be tested and inspected in accordance with part 61, subpart 61.10, of this...

  13. 46 CFR 58.60-3 - Pressure vessel.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 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...

  14. 46 CFR 115.812 - Pressure vessels and boilers.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter....

  15. Neutron flux reduction programs for reactor pressure vessel

    SciTech Connect

    Yoo, C.S.; Kim, B.C.

    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)

  16. Pressure vessel calculations for VVER-440 reactors.

    PubMed

    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.

  17. Midland reactor pressure vessel flaw distribution

    SciTech Connect

    Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

    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. Vehicular Storage of Hydrogen in Insulated Pressure Vessels

    SciTech Connect

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

    2005-01-03

    This paper describes the development of an alternative technology for storing hydrogen fuel onboard automobiles. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept cryogenic liquid fuel, cryogenic compressed gas or compressed gas at ambient temperature. Insulated pressure vessels offer advantages over conventional H{sub 2} storage approaches. Insulated pressure vessels are more compact and require less carbon fiber than GH{sub 2} vessels. They have lower evaporative losses than LH{sub 2} tanks, and are much lighter than metal hydrides. After outlining the advantages of hydrogen fuel and insulated pressure vessels, the paper describes the experimental and analytical work conducted to verify that insulated pressure vessels can be used safely for vehicular H{sub 2} storage. The paper describes tests that have been conducted to evaluate the safety of insulated pressure vessels. Insulated pressure vessels have successfully completed a series of DOT, ISO and SAE certification tests. A draft procedure for insulated pressure vessel certification has been generated to assist in a future commercialization of this technology. An insulated pressure vessel has been installed in a hydrogen fueled truck and it is currently being subjected to extensive testing.

  19. Neural Network Burst Pressure Prediction in Composite Overwrapped Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Hill, Eric v. K.; Dion, Seth-Andrew T.; Karl, Justin O.; Spivey, Nicholas S.; Walker, James L., II

    2007-01-01

    Acoustic emission data were collected during the hydroburst testing of eleven 15 inch diameter filament wound composite overwrapped pressure vessels. A neural network burst pressure prediction was generated from the resulting AE amplitude data. The bottles shared commonality of graphite fiber, epoxy resin, and cure time. Individual bottles varied by cure mode (rotisserie versus static oven curing), types of inflicted damage, temperature of the pressurant, and pressurization scheme. Three categorical variables were selected to represent undamaged bottles, impact damaged bottles, and bottles with lacerated hoop fibers. This categorization along with the removal of the AE data from the disbonding noise between the aluminum liner and the composite overwrap allowed the prediction of burst pressures in all three sets of bottles using a single backpropagation neural network. Here the worst case error was 3.38 percent.

  20. Reactor pressure vessel annealing -- Effective mitigation method

    SciTech Connect

    Brumovsky, M.; Brynda, J.

    1996-09-01

    Reactor pressure vessels of old generation were mostly manufactured from materials with high content of impurities which results in high increase in irradiation embrittlement values. Standard mitigation methods for decrease this damage--application of low-leakage core or dummy elements insertion--are inefficient if applied during the reactor operation. Thermal annealing of reactor pressure vessels has been shown as a very effective method for restoration of initial material properties in a high extent. Even though annealing process is not fully understood from the microstructural changes point of view, results from the testing were so promising that many annealing of WWER RPVs were performed. Nevertheless, some problems still remains, connected mainly with monitoring of the extent of annealing restoration as well as with re-embrittlement rate after such a properties restoration. Experience with WWER-440 RPVs is discussed, mainly because of the austenitic cladding existence. Cladding does not allow to take templates from the inner RPV surface and it is damaged during operation, as well. At the same time, no monitoring of cladding behavior during operation was planned within surveillance programs. Problems connected with material behavior monitoring after annealing as well as during further operation (re-embrittlement rate) are discussed together with the assessment of inaccuracies and possible solutions.

  1. Welded repairs of punctured thin-walled aluminum pressure vessels

    NASA Technical Reports Server (NTRS)

    Jones, D. J.

    1969-01-01

    Punctures in thin-walled aluminum pressure vessels are repaired by plugging the hole with an interference-fit disc and welding the unit. The repaired vessels withstood test pressures in excess of vessel ultimate design values for 2-, 4-, and 6-inch holes in 0.202-inch-thick aluminum alloy parent material.

  2. H.B. Robinson-2 pressure vessel benchmark

    SciTech Connect

    Remec, I.; Kam, F.B.K.

    1998-02-01

    The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

  3. Plating Repair Of Nickel-Alloy Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Ricklefs, Steve K.; Chagnon, Kevin M.

    1989-01-01

    Procedure for localized electrodeposition of nickel enables repair of small damaged nickel-based pressure vessels. Electrodeposition restores weakened areas of vessel wall to at least their former strength.

  4. Chemical Safety Alert: Rupture Hazard of Pressure Vessels

    EPA Pesticide Factsheets

    Pressure vessels or boilers can fail catastrophically if they are not properly designed, constructed, operated, inspected, tested, or repaired. Risk increases if vessels contents are toxic, corrosive, reactive, or flammable.

  5. Acoustic emission testing of 12-nickel maraging steel pressure vessels

    NASA Technical Reports Server (NTRS)

    Dunegan, H. L.

    1973-01-01

    Acoustic emission data were obtained from three point bend fracture toughness specimens of 12-nickel maraging steel, and two pressure vessels of the same material. One of the pressure vessels contained a prefabricated flaw which was extended and sharpened by fatigue cycling. It is shown that the flawed vessel had similar characteristics to the fracture specimens, thereby allowing estimates to be made of its nearness to failure during a proof test. Both the flawed and unflawed pressure vessel survived the proof pressure and 5 cycles to the working pressure, but it was apparent from the acoustic emission response during the proof cycle and the 5 cycles to the working pressure that the flawed vessel was very near failure. The flawed vessel did not survive a second cycle to the proof pressure before failure due to flaw extension through the wall (causing a leak).

  6. Asymmetric Bulkheads for Cylindrical Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Ford, Donald B.

    2007-01-01

    Asymmetric bulkheads are proposed for the ends of vertically oriented cylindrical pressure vessels. These bulkheads, which would feature both convex and concave contours, would offer advantages over purely convex, purely concave, and flat bulkheads (see figure). Intended originally to be applied to large tanks that hold propellant liquids for launching spacecraft, the asymmetric-bulkhead concept may also be attractive for terrestrial pressure vessels for which there are requirements to maximize volumetric and mass efficiencies. A description of the relative advantages and disadvantages of prior symmetric bulkhead configurations is prerequisite to understanding the advantages of the proposed asymmetric configuration: In order to obtain adequate strength, flat bulkheads must be made thicker, relative to concave and convex bulkheads; the difference in thickness is such that, other things being equal, pressure vessels with flat bulkheads must be made heavier than ones with concave or convex bulkheads. Convex bulkhead designs increase overall tank lengths, thereby necessitating additional supporting structure for keeping tanks vertical. Concave bulkhead configurations increase tank lengths and detract from volumetric efficiency, even though they do not necessitate additional supporting structure. The shape of a bulkhead affects the proportion of residual fluid in a tank that is, the portion of fluid that unavoidably remains in the tank during outflow and hence cannot be used. In this regard, a flat bulkhead is disadvantageous in two respects: (1) It lacks a single low point for optimum placement of an outlet and (2) a vortex that forms at the outlet during outflow prevents a relatively large amount of fluid from leaving the tank. A concave bulkhead also lacks a single low point for optimum placement of an outlet. Like purely concave and purely convex bulkhead configurations, the proposed asymmetric bulkhead configurations would be more mass-efficient than is the flat

  7. Comparative analysis of pressure vessel integrity for various LOCA conditions

    NASA Astrophysics Data System (ADS)

    Çolak, Üner; Özdere, Oya

    2001-09-01

    In this study, integrity analysis is performed for a classical four loop PWR pressure vessel fabricated from SA533B type ferritic steel. Pressure vessel behavior is analyzed by deterministic and probabilistic methods under transient conditions, which may cause pressurized thermal shock (PTS). In deterministic analysis, the change of material properties and the mechanical state of the vessel are analyzed against changes in coolant pressure and temperature. Probabilistic analysis is performed to obtain pressure vessel beltline region weld failure probabilities in transient conditions. Overall vessel failure probabilities are evaluated based on the results of deterministic analyses. Computer code VISA-II is utilized for the calculation of vessel failure probabilities. Among three cases considered in this study, a medium break loss of coolant accident induced by a 50 cm2 break in the hot leg yields the highest vessel rupture probability. The maximum nil ductility temperature in all cases is still below the NRC PTS limit.

  8. Reactor pressure vessel head vents and methods of using the same

    SciTech Connect

    Gels, John L; Keck, David J; Deaver, Gerald A

    2014-10-28

    Internal head vents are usable in nuclear reactors and include piping inside of the reactor pressure vessel with a vent in the reactor upper head. Piping extends downward from the upper head and passes outside of the reactor to permit the gas to escape or be forcibly vented outside of the reactor without external piping on the upper head. The piping may include upper and lowers section that removably mate where the upper head joins to the reactor pressure vessel. The removable mating may include a compressible bellows and corresponding funnel. The piping is fabricated of nuclear-reactor-safe materials, including carbon steel, stainless steel, and/or a Ni--Cr--Fe alloy. Methods install an internal head vent in a nuclear reactor by securing piping to an internal surface of an upper head of the nuclear reactor and/or securing piping to an internal surface of a reactor pressure vessel.

  9. Auger spectroscopy of Magnox pressure vessel steels

    SciTech Connect

    Fisher, S.; Knowles, G.; Lee, B.

    1999-10-01

    Magnox Electric maintains a significant microstructural program in support of its safety case for operation of its stations with steel pressure vessels. An important part of this program is the characterization of grain boundary chemistry using Auger spectroscopy. Mechanical testing and subsequent examination of surveillance material has shown that some Charpy specimens display a proportion of intergranular fracture and Auger work has linked this to the presence of phosphorus on the grain boundaries. A feature of particular interest in the study of the boundaries is the co-segregation of carbon. The measurement of the true levels of phosphorus and carbon segregation is complicated by the presence of carbon contamination. This paper describes the simple approach used to overcome this problem.

  10. Environmental Testing of Glass-Fiber/Epoxy Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Faddoul, J. R.

    1987-01-01

    Pair of reports discusses long-term environmental tests of glassfiber/epoxy composite pressure vessels. Strength diminishes during long exposure to environment. Since such data necessary for accurate design of long-life structures such as pressure vessels, NASA Lewis Research Center built outdoor test stand in 1973. Test stand maintains system under constant pressure loading without frequent intervention of personnel.

  11. High-pressure cryogenic seals for pressure vessels

    NASA Technical Reports Server (NTRS)

    Buggle, A. E.

    1977-01-01

    Problems associated with maintaining high pressures at cryogenic temperatures in pressure vessels are investigated. The goals were to identify the appropriate materials and design for a seal intended for cryogenic applications at pressures up to 4,080 bars (60,000 psi), and to examine the factors affecting the seal performance. The method employed and the apparatus used in a series of experimental seal system tests, and the test results are described in detail. It is concluded that the common seal designs and extrusion seal-ring materials such as Teflon, tin, and lead are not suitable. However, new seal systems developed using indium seal rings, brass or 304 stainless steel anvil rings, and two 0-rings of silicone rubber or Kel-F did prove suitable.

  12. Holographic and acoustic emission evaluation of pressure vessels

    SciTech Connect

    Boyd, D.M.

    1980-03-05

    Optical holographic interfereometry and acoustic emission monitoring were simultaneously used to evaluate two small, high pressure vessels during pressurization. The techniques provide pressure vessel designers with both quantitative information such as displacement/strain measurements and qualitative information such as flaw detection. The data from the holographic interferograms were analyzed for strain profiles. The acoustic emission signals were monitored for crack growth and vessel quality.

  13. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    SciTech Connect

    Cooke, Conrad; Spann, Holger

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle Plates to

  14. Buffered explosions in steel pressure vessels

    SciTech Connect

    Glenn, L.A.

    1986-01-01

    The impulse delivered to the walls of a vessel containing an explosion will increase if material is placed between the walls and the charge. If the impulse application time is small in compared with the eigenperiod of the vessel, the wall stress will increase in direct proportion to the impulse. Conversely, if the application period can be extended beyond half the eigenperiod, the peak stress will be proportional to the ratio of the impulse to the delivery period. With powder or granular buffers, it is possible for the delivery period to increase faster than the impulse as the buffer mass is increased. This is the reason why certain powders, or porous materials, can provide stress reduction even below that observed by evacuating the space between the walls and the explosive. If the buffer material is to serve as an effective mitigator, it must collapse on shock loading to a final density that depends only weakly on pressure; the criterion is that the wave speed in the material that impacts the wall must be small comparison with the impact (particle) speed. This behavior apparently occurs with salt, at least for modest values of the charge parameter, but to a lesser extent with snow under the same conditions. The vermiculite data are comparable to the salt in the charge paramete region where the two overlap; with increasing explosive, however, the vermiculite appears to behave like the snow and its effectiveness as a mitigator rapidly diminishes. It is also clear that once the wave speed criterion is seriously violated, the use of a powder buffer will result in a higher wall stress than if only air filled the space between walls and charge. 5 refs.

  15. Fracture mechanics toughness behavior of pressure vessel steels in the ductile-to-brittle transition region: An important issue to nuclear reactor integrity

    SciTech Connect

    DeAquino, C.T.; Andrade, A.H.P.; Liendo, M.F.; Landes, J.D.; McCabe, D.E.

    1996-12-01

    ASTM E-08 Committee has been developing a new standard, to deal with the fracture mechanics behavior of steels in the ductile to brittle transition region. This paper presents a comparison between the current approach and a new proposal to be used by the nuclear industry to face the problem of determining the behavior of ferritic steels. An emphasis will be given to the application of this proposal and its evaluation using a Brazilian A508 Class 3 nuclear steel.

  16. Pressurized thermal shock probabilistic fracture mechanics sensitivity analysis for Yankee Rowe reactor pressure vessel

    SciTech Connect

    Dickson, T.L.; Cheverton, R.D.; Bryson, J.W.; Bass, B.R.; Shum, D.K.M.; Keeney, J.A.

    1993-08-01

    The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods used in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117.

  17. Pressure vessel burst test program - Initial program paper

    NASA Technical Reports Server (NTRS)

    Cain, Maurice R.; Sharp, Douglas E.; Coleman, Michael D.; Webb, Bobby L.

    1990-01-01

    The current status of a pressure vessel burst test program, aimed at the study of the blast waves and fragmentation characteristics of ruptured gas-filled pressure vessels, is reported. The program includes a series of test plans, each involving multiple bursts with burst pressures ranging to 7500 psig. The discussion covers the identification of concerns and hazards, application of the data generated, and a brief review of the current methods for assessing vessel safety and burst parameters. Attention is also given to pretest activities, including completed vessel and facility/instrumentation preparation and results of completed preliminary burst tests.

  18. Reactor Pressure Vessel Head Packaging & Disposal

    SciTech Connect

    Wheeler, D. M.; Posivak, E.; Freitag, A.; Geddes, B.

    2003-02-26

    Reactor Pressure Vessel (RPV) Head replacements have come to the forefront due to erosion/corrosion and wastage problems resulting from the susceptibility of the RPV Head alloy steel material to water/boric acid corrosion from reactor coolant leakage through the various RPV Head penetrations. A case in point is the recent Davis-Besse RPV Head project, where detailed inspections in early 2002 revealed significant wastage of head material adjacent to one of the Control Rod Drive Mechanism (CRDM) nozzles. In lieu of making ASME weld repairs to the damaged head, Davis-Besse made the decision to replace the RPV Head. The decision was made on the basis that the required weld repair would be too extensive and almost impractical. This paper presents the packaging, transport, and disposal considerations for the damaged Davis-Besse RPV Head. It addresses the requirements necessary to meet Davis Besse needs, as well as the regulatory criteria, for shipping and burial of the head. It focuses on the radiological characterization, shipping/disposal package design, site preparation and packaging, and the transportation and emergency response plans that were developed for the Davis-Besse RPV Head project.

  19. Thermodynamics of insulated pressure vessels for vehicular hydrogen storage

    SciTech Connect

    Aceves, S.M.; Berry, G.D.

    1997-06-01

    This paper studies the application of insulated pressure vessels for hydrogen-fueled light-duty vehicles. Insulated pressure vessels can store liquid hydrogen (LH2); low-temperature (90 K) compressed hydrogen (CH2); or ambient temperature CH2. In this analysis, hydrogen temperatures, pressures and venting losses am calculated for insulated pressure vessels fueled with LH2 or with low-temperature CH2, and the results are compared to those obtained in low-pressure LH2 tanks. Hydrogen losses are calculated as a function of daily driving distance during normal operation; as a function of time during long periods of vehicle inactivity; and as a function of initial vessel temperature during fueling. The number of days before any venting losses occur is also calculated as a function of the daily driving distance. The results show that insulated pressure vessels have packaging characteristics comparable to those of conventional, low-pressure LH2 tanks (low weight and volume), with greatly improved dormancy and much lower boil-off. Insulated pressure vessels used in a 17 km/l (40 mpg) car do not lose any hydrogen when the car is driven at least 15 km/day in average. Since almost all cars are driven for longer distances, most cars would never lose any hydrogen. Losses during long periods of parking are also relatively small. Due to their high-pressure capacity, these vessels would retain about a third of their full charge even after a very long dormancy, so that the owner would not risk running out of fuel. If an insulated pressure vessel reaches ambient temperature, it can be cooled down very effectively by fueling it with LH2 with no losses during fueling. The vessel has good thermal performance even when thermally insulated with inexpensive microsphere insulation. In addition, the insulated pressure vessels greatly ease fuel availability and infrastructure requirements, since it would be compatible with both compressed and cryogenic hydrogen reveling.

  20. Corrosion of steel tendons used in prestressed concrete pressure vessels

    SciTech Connect

    Griess, J.C.; Naus, D.J.

    1980-01-01

    The corrosion behavior of a high-strength steel (Specifications for Uncoated Seven-Wire-Stress-Relieved Strand for Prestressed Concrete (ASTM A 416-74, Grade 270)), typical of those used as tensioning tendons in prestressed concrete pressure vessels was measured in several corrosive environments. The protection obtained by coating the steel with two commercial petroleum-base greases or with Portland cement grout was evaluated. The few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors were reviewed. The susceptibility of the steel to stress corrosion cracking and hydrogen embrittlement and its general corrosion rate were determined in several salt solutions. Wires coated with the greases and grout were soaked for long periods in the same solutions and changes in their mechanical properties were subsequently determined. All three coatings appeared to give essentially complete protection; however, flaws in the grease coatings could be detrimental, and flaws or cracks less than 1-mm-wide (0.04 in.) in the grout were without effect.

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

    PubMed

    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.

  2. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    SciTech Connect

    GRIFFIN, PATRICK J.

    1999-09-14

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation.

  3. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John

    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.

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

    SciTech Connect

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    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.

  5. Fatigue behavior of reactor pressure vessel steels

    SciTech Connect

    Huang, J.Y.; Chen, C.Y.; Chien, K.F.; Kuo, R.C.; Liaw, P.K.; Huang, J.G.

    1999-07-01

    High-cycle fatigue tests have been conducted on reactor pressure vessel steels, SA533-B1, with four levels of sulfur contents at room temperature. The applied stress versus fatigue life cycle (S-N) curves were developed at load ratios, R, of 0.2 and 0.8. At a load ratio of 0.2, the fatigue limit for SA533-B1 steels with sulfur contents less than 0.015 wt % is around 650 MPa, which is slightly higher than that with sulfur contents higher than 0.027 wt %. At a load ratio of 0.8, there were no fatigue indications on the fracture surface. In some fatigue-tested specimens, specifically those with higher sulfur content levels, fatigue cracks were observed to initiate around the inclusions. A digital video camera was used to record the entire fatigue process, and the results demonstrated that the crack initiation period dominated more than 80% of the total fatigue life. The fatigue-tested specimen surface had been thoroughly examined using optical and scanning electron microscopy. Apparent distinctions were observed between the neighborhood of the crack initiation site and the rest of the specimen surface. A great number of precipitates were found distributed along the sub-grain boundary using transmission electron microscopy. There is no or little change of the morphology of precipitates before and after fatigue tests. The mis-orientation between two neighboring sub-grains ranges from 1 to 5{degree}. The effects of the applied maximum stress, precipitate distribution, and fatigue cycle on the mis-orientation of the sub-grain boundary will be discussed in this paper.

  6. Removal of the Yankee pressure vessel: Diary of a work in progress

    SciTech Connect

    Child, C.L.

    1996-09-01

    On February 26, 1992, The Yankee Atomic Electric Company (YAEC) announced the permanent shut down of the Yankee Nuclear Power Station (YNPS). As part of decommissioning and since early 1995, Yankee has embarked on a program to remove, transport and dispose of the reactor pressure vessel. This paper is a report of the progress to date, and future activities. Actual removal of the reactor pressure vessel will proceed in the reverse order in which it was installed. The upper neutron shield tank will be removed down to the reactor support ring, exposing both the reactor coolant piping and the upper insulation about the reactor pressure vessel and piping. The insulation, which contains asbestos, will be removed as will the reactor coolant piping between the reactor pressure vessel nozzles and the Reactor Support Structure interior wall. The reactor pressure vessel will be lifted from the bottom of the shield tank cavity and placed within a cask which has been oriented below the vapor container equipment hatch. Once in the cask, concrete will be placed both within the reactor pressure vessel and between the reactor pressure vessel and the cask to fix any loose contamination, and to provide additional shielding. Following curing of the concrete, the cask will be down ended to a horizontal position and prepared for transport and disposal. The reactor pressure vessel cask package will be transported overland to the nearest rail line, and will proceed by rail to a disposal site. Reactor pressure vessel lift and transport, originally scheduled for the Spring of 1996, will be delayed until the NRC reapproves the YNPS Decommissioning Plan.

  7. High-performance fiber/epoxy composite pressure vessels

    NASA Technical Reports Server (NTRS)

    Chiao, T. T.; Hamstad, M. A.; Jessop, E. S.; Toland, R. H.

    1978-01-01

    Activities described include: (1) determining the applicability of an ultrahigh-strength graphite fiber to composite pressure vessels; (2) defining the fatigue performance of thin-titanium-lined, high-strength graphite/epoxy pressure vessel; (3) selecting epoxy resin systems suitable for filament winding; (4) studying the fatigue life potential of Kevlar 49/epoxy pressure vessels; and (5) developing polymer liners for composite pressure vessels. Kevlar 49/epoxy and graphite fiber/epoxy pressure vessels, 10.2 cm in diameter, some with aluminum liners and some with alternation layers of rubber and polymer were fabricated. To determine liner performance, vessels were subjected to gas permeation tests, fatigue cycling, and burst tests, measuring composite performance, fatigue life, and leak rates. Both the metal and the rubber/polymer liner performed well. Proportionately larger pressure vessels (20.3 and 38 cm in diameter) were made and subjected to the same tests. In these larger vessels, line leakage problems with both liners developed the causes of the leaks were identified and some solutions to such liner problems are recommended.

  8. Evaluation of insulated pressure vessels for cryogenic hydrogen storage

    SciTech Connect

    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

    NASA Technical Reports Server (NTRS)

    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. Common/Dependent-Pressure-Vessel Nickel-Hydrogen Batteries

    NASA Technical Reports Server (NTRS)

    Timmerman, Paul J.

    2003-01-01

    The term "common/dependent pressure vessel" (C/DPV) denotes a proposed alternative configuration for a nickelhydrogen battery. The C/DPV configuration is so named because it is a hybrid of two prior configurations called "common pressure vessel" (CPV) and "dependent pressure vessel" (DPV). The C/DPV configuration has been proposed as a basis for designing highly reliable, long-life Ni/H2-batteries and cells for anticipated special applications in which it is expected that small charge capacities will suffice and sizes and weights must be minimized.

  11. On the optimal pretensioning of cylindrical and spherical pressure vessels

    SciTech Connect

    Kalamkarov, A.L.; Drozdov, A.D.

    1995-11-01

    Filament winding of pressure vessels and pipes is always realized with some pretensioning, and some external loads may be applied. It is important to determine such an optimal preload regime that ensures the maximum load-carrying capacity of the vessel subject to internal pressure. In the present study, the optimal preload distribution is analyzed in the filament winding fabrication of the cylindrical or spherical pressure vessels that are treated as growing elastic solids subjected to aging. In the case of cylindrical vessels, the dependence of the optimal preload intensity versus the polar radius is obtained for both nonaging and aging material of the fibers. In the case of spherical pressure vessels, the optimal regime of internal pressure applied during the winding process is obtained. The optimal loading of a spherical vessel at both infinitesimal and finite strains is analyzed. The new solutions obtained and the recommendations formulated are of a special practical importance for the optimal design and fabrication of the composite pressure vessels and pipes.

  12. Quantification of Processing Effects on Filament Wound Pressure Vessels

    NASA Technical Reports Server (NTRS)

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

  14. Lightweight cryogenic-compatible pressure vessels for vehicular fuel storage

    DOEpatents

    Aceves, Salvador; Berry, Gene; Weisberg, Andrew H.

    2004-03-23

    A lightweight, cryogenic-compatible pressure vessel for flexibly storing cryogenic liquid fuels or compressed gas fuels at cryogenic or ambient temperatures. The pressure vessel has an inner pressure container enclosing a fuel storage volume, an outer container surrounding the inner pressure container to form an evacuated space therebetween, and a thermal insulator surrounding the inner pressure container in the evacuated space to inhibit heat transfer. Additionally, vacuum loss from fuel permeation is substantially inhibited in the evacuated space by, for example, lining the container liner with a layer of fuel-impermeable material, capturing the permeated fuel in the evacuated space, or purging the permeated fuel from the evacuated space.

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

    SciTech Connect

    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.

  16. Crashworthy sealed pressure vessel for plutonium transport

    SciTech Connect

    Andersen, J.A.

    1980-01-01

    A rugged transportation package for the air shipment of radioisotopic materials was recently developed. This package includes a tough, sealed, stainless steel inner containment vessel of 1460 cc capacity. This vessel, intended for a mass load of up to 2 Kg PuO/sub 2/ in various isotopic forms (not to exceed 25 watts thermal activity), has a positive closure design consisting of a recessed, shouldered lid fastened to the vessel body by twelve stainless-steel bolts; sealing is accomplished by a ductile copper gasket in conjunction with knife-edge sealing beads on both the body and lid. Follow-on applications of this seal in newer, smaller packages for international air shipments of plutonium safeguards samples, and in newer, more optimized packages for greater payload and improved efficiency and utility, are briefly presented.

  17. SOLID GLASS AND CERAMIC EXTERNAL-PRESSURE VESSELS

    DTIC Science & Technology

    characteristics and withstands external pressure cycling and mild underwater dynamic pressures. Scratches on the exterior surfaces do not decrease...appreciably the compressive and elastic strength of such vessels when exposed to either static or cycling pressure. Connectors have been devised that enable

  18. Proceedings of the 1985 pressure vessels and piping conference. Volume PVP-98-2. Pressure vessel components design and analysis

    SciTech Connect

    Gawaltney, R.C.

    1985-01-01

    There are seven sessions covered in this book on Pressure Vessel Components Design and Analysis. The papers are divided into the following six subject areas: composites, valves, tubesheets, pressure vessels and piping, bolted flanges, and nonlinear computational methods. The design procedures and analysis methods described in this book are not discussed previously. The engineers working in the field of pressure vessel design can only keep up with current developments in these areas by reviewing a substantial amount of technical literature. A goal of this book is to help in this endeavor by offering selected papers in the area by authors who are experienced and distinguished workers in their fields.

  19. Thick-wall Kevlar 49/Epoxy pressure vessels

    SciTech Connect

    Guess, T.R.

    1984-01-01

    The feasibility of thick-wall composite vessels for very high pressure applications is demonstrated. Prototype vessels, in both spherical and cylindrical geometries, were designed, fabricated and burst tested. It is shown that experimental burst pressures are in excellent agreement with predicted values for burst pressures up to 60 ksi. Each unit consisted of a thin, seamless, copper liner with stainless steel fill stems and a filament-wound Kevlar 49/epoxy outer shell. Analysis of vessel performance accounted for liner thickness and yield strengths, composite thickness, mechanical properties and fiber volume fraction, and stress concentrations caused by the fill stem. Spherical vessels of three different sizes (inside diameters of 2.15 inches, 4.0 inches and 5.3 inches) with either 30 ksi or 60 ksi design burst pressure are discussed. Also, cylindrical vessels with identical liners but of two different composite thicknesses are described. These vessels achieved 50 ksi and 57 ksi burst pressures, respectively. In addition to the design considerations alluded to throughout the paper, the stress state in a thin metal liner during cyclic loading and the life prediction of composite vessels under sustained loading are discussed.

  20. Float level switch for a nuclear power plant containment vessel

    DOEpatents

    Powell, J.G.

    1993-11-16

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures.

  1. Float level switch for a nuclear power plant containment vessel

    DOEpatents

    Powell, James G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  2. Stiffness Study of Wound-Filament Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Verderaime, V.

    1986-01-01

    Report presents theoretical and experimental study of stiffness of lightweight, jointed pressure vessels made of wound graphite fibers and epoxy. Specimens fabricated from layers of graphite fibers, wet with epoxy, on aluminum mandrel. Segment ends thickened with interspersed layers of axially oriented fibers to reduce pinhole bearing stresses and local deformations. Segments cured at 390 degrees F (199 degrees C). Report presents results of vibrational tests of one-quarter-scale models of wound-filament pressure vessels.

  3. Creep of A508/533 Pressure Vessel Steel

    SciTech Connect

    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

  4. The behavior of shallow flaws in reactor pressure vessels

    SciTech Connect

    Rolfe, S.T. )

    1991-11-01

    Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs.

  5. Progressive Fracture and Damage Tolerance of Composite Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Chamis, Christos C.; Gotsis, Pascal K.; Minnetyan, Levon

    1997-01-01

    Structural performance (integrity, durability and damage tolerance) of fiber reinforced composite pressure vessels, designed for pressured shelters for planetary exploration, is investigated via computational simulation. An integrated computer code is utilized for the simulation of damage initiation, growth, and propagation under pressure. Aramid fibers are considered in a rubbery polymer matrix for the composite system. Effects of fiber orientation and fabrication defect/accidental damages are investigated with regard to the safety and durability of the shelter. Results show the viability of fiber reinforced pressure vessels as damage tolerant shelters for planetary colonization.

  6. Dimensional analysis of blood vessels in the pressure myograph

    NASA Astrophysics Data System (ADS)

    Crabtree, Vincent P.; Smith, Peter R.

    1999-01-01

    The accuracy of conventional and emerging methods for the dimensional analysis of optically imaged arterial vessels, isolated in a pressure myograph, is investigated. The pressure myograph is a device used to study the structure and function of isolated sections of small resistance arteries, as a function of chemical, mechanical and electrical stimuli. The arterial wall and lumen dimensions are particularly important indicators of anatomy and pathology. The conventional method of dimensional analysis uses edge detection, however the accuracy of this approach is questionable when the vessel is in a contracted state since contrast deteriorates or is lost between lumen and vessel wall. The conventional and emerging methods are examined experimentally with vessel phantoms, to provide known characteristics. A novel algorithm, based on a measurement of the vessel extinction coefficient, is also examined theoretically and experimentally. A discussion centers on the possibility for realistic lumen size measurement when edge detection can not be applied and when the accuracy of edge detection is questionable.

  7. Low-Cost, Lightweight Pressure Vessel Proof Test

    NASA Astrophysics Data System (ADS)

    Chanez, Eric

    This experiment seeks to determine the burst strength of the low-cost, lightweight pressure vessel fabricated by the Suborbital Center of Excellence (SCE). Moreover, the test explores the effects of relatively large gage pressures on material strain for ‘pumpkin-shaped' pressure vessels. The SCE team used pressure transducers and analog gauges to measure the gage pressure while a video camera assembly recorded several gores in the shell for strain analysis. The team loaded the vessel in small intervals of pressure until the structure failed. Upon test completion, the pressure readings and video recordings were analyzed to determine the burst strength and material strain in the shell. The analysis yielded a burst pressure of 13.5 psi while the strain analysis reported in the shell. While the results of this proof test are encouraging, the structure's factor of safety must be increased for actual balloon flights. Furthermore, the pressure vessel prototype must be subjected to reliability tests to show the design can sustain gage pressures for the length of a balloon flight.

  8. Time-dependent response of filamentary composite spherical pressure vessels

    NASA Technical Reports Server (NTRS)

    Dozier, J. D.

    1983-01-01

    A filamentary composite spherical pressure vessel is modeled as a pseudoisotropic (or transversely isotropic) composite shell, with the effects of the liner and fill tubes omitted. Equations of elasticity, macromechanical and micromechanical formulations, and laminate properties are derived for the application of an internally pressured spherical composite vessel. Viscoelastic properties for the composite matrix are used to characterize time-dependent behavior. Using the maximum strain theory of failure, burst pressure and critical strain equations are formulated, solved in the Laplace domain with an associated elastic solution, and inverted back into the time domain using the method of collocation. Viscoelastic properties of HBFR-55 resin are experimentally determined and a Kevlar/HBFR-55 system is evaluated with a FORTRAN program. The computed reduction in burst pressure with respect to time indicates that the analysis employed may be used to predict the time-dependent response of a filamentary composite spherical pressure vessel.

  9. Stress analysis and evaluation of a rectangular pressure vessel

    NASA Astrophysics Data System (ADS)

    Rezvani, M. A.; Ziada, H. H.; Shurrab, M. S.

    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 8; 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 8, division 1 instead of division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel.

  10. Advanced composite fiber/metal pressure vessels for aircraft applications

    NASA Astrophysics Data System (ADS)

    Papanicolopoulos, Aleck

    1993-06-01

    Structural Composites Industries has developed, qualified, and delivered a number of high performance carbon epoxy overwrapped/seamless aluminum liner pressure vessels for use in military aircraft where low weight, low cost, high operating pressure and short lead time are the primary considerations. This paper describes product design, development, and qualification for a typical program. The vessel requirements included a munitions insensitivity criterion as evidenced by no fragmentation following impact by a .50 cal tumbling bullet. This was met by the development of a carbon-Spectra hybrid composite overwrap on a thin-walled seamless aluminum liner. The same manufacturing, inspection, and test processes that are used to produce lightweight, thin walled seamless aluminum lined carbon/epoxy overwrapped pressure vessels for satellite and other space applications were used to fabricate this vessel. This report focuses on the results of performance in the qualification testing.

  11. Light Water Reactor-Pressure Vessel Surveillance project computer system

    SciTech Connect

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes.

  12. Integrity of PWR pressure vessels during overcooling accidents

    SciTech Connect

    Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation.

  13. Radiation embrittlement and annealing of VVER pressure vessels

    SciTech Connect

    Weeks, J.R. )

    1989-01-01

    Pressure vessels of the Soviet-designed VVERs are exposed to up to 10X the fast flux of the vessels of US PWRs. They are fabricated of 2 1/2% Cr, 1% Mo or 2 1/2% Cr, 1% Ni steels developed for that purpose. Consequently, the data base on irradiation effects differs somewhat from that of Western-designed pressure vessels. The role of phosphorus, which is high in the older VVER steels, is especially important. Newer grades of steels, low in copper and phosphorus, have been developed. Pressure vessels fabricated before about 1980 were unclad. The embrittlement of these can be tested in situ with a remote microhardness measuring device. Hydrogen pickup from corrosion does not increase the embrittlement due to irradiation. Considerable research has been performed on pressure vessel annealing, and anneals of the core weld region of the pressure vessels of three operating VVERs have been completed at temperatures 160-185{degree}C higher than the irradiation temperature for 150 hours to one week. Recoveries, monitored with a remote microhardness tester, have ranged upward from 70%.

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

    SciTech Connect

    Corwin, W.R.; Broadhead, B.L.; Suzuki, M.; Kohsaka, A.

    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).

  15. Validation and Verification of Composite Pressure Vessel Design

    NASA Technical Reports Server (NTRS)

    Kreger, Stephen T.; Ortyl, Nicholas; Grant, Joseph; Taylor, F. Tad

    2006-01-01

    Ten composite pressure vessels were instrumented with fiber Bragg grating sensors and pressure tested Through burst. This paper and presentation will discuss the testing methodology, the test results, compare the testing results to the analytical model, and also compare the fiber Bragg grating sensor data with data obtained against that obtained from foil strain gages.

  16. Surveillance of WWER-440C reactor pressure vessels

    SciTech Connect

    Brumovsky, M.; Pav, T.

    1993-12-01

    In Czechoslovakia there are six units of Water-Water Power Reactor (WWER)-440 C type reactors (pressurized water reactor [PWR] type) incorporated with pressure vessel surveillance specimens. These sets of specimens are kept for carrying out static tensile testing, impact notch toughness testing, and static fracture toughness testing, and are supplemented by necessary sets of neutron flux monitors. Results of mechanical testing of these specimens evaluated after one to five years of reactor operation are summarized and discussed with respect to the effect of individual heats and welding joints, radiation embrittlement laws, and lead factor and pressure vessel lifetime assessment.

  17. Advanced technology for minimum weight pressure vessel system

    NASA Technical Reports Server (NTRS)

    Hamstad, M. A.; Jessop, E. S.; Toland, R. H.

    1977-01-01

    Bosses were made of fiber/resin composite materials to evaluate their potential in lightweight pressure vessels. An approximate 25% weight savings over the standard aluminum boss was achieved without boss failures during burst tests. Polymer liners and metal liners are used in fiber composite pressure vessels for containment of gases. The internal support of these liners required during the filament winding process has previously been provided by dissolvable salt mandrels. An internal pressurization technique has been developed which allows overwinding the liner without other means of support and without collapse. Study was made of several additional concepts including styrene/Saran, styrene/flexible epoxy.

  18. Optimization of multilayered composite pressure vessels using exact elasticity solution

    SciTech Connect

    Adali, S.; Verijenko, V.E.; Tabakov, P.Y.; Walker, M.

    1995-11-01

    An approach for the optimal design of thick laminated cylindrical pressure vessels is given. The maximum burst pressure is computed using an exact elasticity solution and subject to the Tsai-Wu failure criterion. The design method is based on an accurate 3-D stress analysis. Exact elasticity solutions are obtained using the stress function approach where the radial, circumferential and shear stresses are determined taking the closed ends of the cylindrical shell into account. Design optimization of multilayered composite pressure vessels are based on the use of robust multidimensional methods which give fast convergence. Two methods are used to determine the optimum ply angles, namely, iterative and gradient methods. Numerical results are given for optimum fiber orientation of each layer for thick and thin-walled multilayered pressure vessels.

  19. Transient Response of FGM Pressure Vessels

    NASA Astrophysics Data System (ADS)

    Pekel, Hakan; Keles, Ibrahim; Temel, Beytullah; Tutuncu, Naki

    The present study aims to investigate the transient behavior of thick-walled cylinders under dynamic internal pressure. Analytical solutions are possible only for simple time-dependent pressure functions. The solution procedure presented is general in the sense that the pressure applied may be an arbitrary continuous function of time, impulsive or given in a discrete form. The material considered is isotropic and heterogeneous with properties varying in the radial direction termed as Functionally Graded Material (FGM). Laplace transform method is used and the inversion into the time domain is performed using the modified Durbin's method. Verification of the numerical procedure is performed by comparing the results with those of an analytical solution available in the literature for a simple exponentially-varying pressure. The inhomogeneity constant in the material property model is shown to have a significant effect on the transient response.

  20. Composite Overwrapped Pressure Vessel (COPV) Stress Rupture Testing

    NASA Technical Reports Server (NTRS)

    Greene, Nathanael J.; Saulsberry, Regor L.; Leifeste, Mark R.; Yoder, Tommy B.; Keddy, Chris P.; Forth, Scott C.; Russell, Rick W.

    2010-01-01

    This paper reports stress rupture testing of Kevlar(TradeMark) composite overwrapped pressure vessels (COPVs) at NASA White Sands Test Facility. This 6-year test program was part of the larger effort to predict and extend the lifetime of flight vessels. Tests were performed to characterize control parameters for stress rupture testing, and vessel life was predicted by statistical modeling. One highly instrumented 102-cm (40-in.) diameter Kevlar(TradeMark) COPV was tested to failure (burst) as a single-point model verification. Significant data were generated that will enhance development of improved NDE methods and predictive modeling techniques, and thus better address stress rupture and other composite durability concerns that affect pressure vessel safety, reliability and mission assurance.

  1. Composite Overwrapped Pressure Vessel(COPV) Stress Rupture Testing

    NASA Astrophysics Data System (ADS)

    Greene, Nathanael J.; Saulsberry, Regor L.; Leifeste, Mark, R.; Yoder, Tommy B.; Keddy, Chris P.; Forth, Scott C.; Russell, Rick W.

    2010-09-01

    This paper reports stress rupture testing of Kevlar® composite overwrapped pressure vessels(COPVs) at NASA White Sands Test Facility. This 6-year test program was part of the larger effort to predict and extend the lifetime of flight vessels. Tests were performed to characterize control parameters for stress rupture testing, and vessel life was predicted by statistical modeling. One highly instrumented 102-cm(40-in.) diameter Kevlar® COPV was tested to failure(burst) as a single-point model verification. Significant data were generated that will enhance development of improved NDE methods and predictive modeling techniques, and thus better address stress rupture and other composite durability concerns that affect pressure vessel safety, reliability and mission assurance.

  2. Nickel hydrogen multicell common pressure vessel battery development update

    NASA Technical Reports Server (NTRS)

    Zagrodnik, Jeffrey P.; Jones, Kenneth R.

    1992-01-01

    The technology background and design qualification of the multicell common pressure vessel nickel hydrogen battery are described. The results of full flight qualification, including random vibration at 19.5 g for two minutes in each axis, electrical characterization in a thermal vacuum chamber, and mass spectroscopy vessel leak detection are reviewed and 12.7 cm qualification and 25.4 cm design adaptation are discussed.

  3. `Sausage string' patterns in blood vessels at high blood pressures

    NASA Astrophysics Data System (ADS)

    Alstrøm, Preben; Eguíluz, Victor M.; Gustafsson, Finn; Holstein-Rathlou, Niels-Henrik

    A new Rayleigh-type instability is proposed to explain the `sausage-string' pattern of alternating constrictions and dialtations formed in blood vessels at high blood pressure conditions. Our theory involves the nonlinear stress-strain characteristics of the vessel wall, and provides predictions for the conditions under which the normal cylindrical geometry of a blood vessel becomes unstable. The theory explains key features observed experimentally, e.g. the limited occurrence of the sausage-string pattern to small arteries and large arterioles, and only in those with small wall-to-lumen ratios.

  4. Assessment of uncertainties for PWR pressure vessel surveillance-French Experience

    SciTech Connect

    Kodeli, I.; Nimal, J.C.

    1996-12-31

    A characteristic of the French nuclear installations that differs from those in most other countries with an important nuclear industry is their high degree of standardization. Two main types of pressurized water reactors (PWRs) are the 900-MW(electric) CPY and the 1300-MW(electric) P4 reactors produced by a single manufacturer, Electricite de France (EdF). Loading schemes are very standardized, although greater diversification has been introduced in recent years due to implementation of some new loading schemes (LLLP, mixed oxide, extended fuel cycle). This report describes the pressure vessel surveillance program.

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

    SciTech Connect

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1995-12-31

    The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is i good surrogate for shift recovery and that there is a high level of consistency between he observed annealing trends and fundamental models of embrittlement and recovery processes.

  6. Advances in crack-arrest technology for reactor pressure vessels

    SciTech Connect

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs.

  7. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except...

  8. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 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...

  9. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Code of Federal Regulations, 2010 CFR

    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...

  10. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 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 Operations...

  11. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 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 Operations...

  12. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 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 Operations...

  13. Pressure vessel burst test program - Progress paper No. 3

    NASA Technical Reports Server (NTRS)

    Cain, Maurice R.; Sharp, Douglas E.; Coleman, Michael D.

    1992-01-01

    An updated progress report is provided on a program developed to study through test and analysis, the characteristics of blast waves and fragmentation generated by ruptured gas filled pressure vessels. Prior papers on this USAF/NASA/General Physics program were presented to the AIAA in July 1990 and June 1991. Ten pressure vessels have been burst using pneumatic pressure. Tests were designed to explore burst characteristics and used an instrumented arena. Data trends for current experiments are presented. This paper is the third progress report on the program and addresses: (1) a brief review of current methods for assessing vessel safety and burst parameters, (2) a review of pneumatic burst testing operations and testing results, including a comparison to current methods for burst assessment, and (3) a review of the basis for the current test program including planned testing.

  14. Beryllium pressure vessels for creep tests in magnetic fusion energy

    SciTech Connect

    Neef, W.S.

    1990-07-20

    Beryllium has interesting applications in magnetic fusion experimental machines and future power-producing fusion reactors. Chief among the properties of beryllium that make these applications possible is its ability to act as a neutron multiplier, thereby increasing the tritium breeding ability of energy conversion blankets. Another property, the behavior of beryllium in a 14-MeV neutron environment, has not been fully investigated, nor has the creep behavior of beryllium been studied in an energetic neutron flux at thermodynamically interesting temperatures. This small beryllium pressure vessel could be charged with gas to test pressures around 3, 000 psi to produce stress in the metal of 15,000 to 20,000 psi. Such stress levels are typical of those that might be reached in fusion blanket applications of beryllium. After contacting R. Powell at HEDL about including some of the pressure vessels in future test programs, we sent one sample pressure vessel with a pressurizing tube attached (Fig. 1) for burst tests so the quality of the diffusion bond joints could be evaluated. The gas used was helium. Unfortunately, budget restrictions did not permit us to proceed in the creep test program. The purpose of this engineering note is to document the lessons learned to date, including photographs of the test pressure vessel that show the tooling necessary to satisfactorily produce the diffusion bonds. This document can serve as a starting point for those engineers who resume this task when funds become available.

  15. Liquid-Level Monitor for Pressurized Vessels

    NASA Technical Reports Server (NTRS)

    Singh, J. J.; Davis, W. T.; Mall, G. H.

    1986-01-01

    Technique for monitoring water levels in pressurized stainless-steel cylinders, based on differences in gamma-ray attenuation coefficients in water and air, developed. Full-scale laboratory prototype system constructed to test technique. Technique usable with liquids other than water, since linear attenuation coefficients for intermediate-energy gamma rays in air considerably lower than in liquids. Also adaptable for continuous monitoring of liquid levels in resevoir systems and in underground storage tanks.

  16. Maine Yankee dosimetry capsule and pressure vessel neutron fluence calculations

    SciTech Connect

    White, J.R.; Spinney, K.B.; Morrissey, K.J.; Cacciapouti, R.J.

    1994-12-31

    In-house capability for deterministic neutron and gamma transport analyses has been implemented at Yankee Atomic Electric Company (YAEC). A detailed R-Theta (R-{theta}) calculational model of Maine Yankee was developed to help in validation of the methods and to establish appropriate models for support of the ongoing Maine Yankee pressure vessel surveillance program. Several data and modeling sensitivity studies were performed and comparisons to measured dosimetry capsule data were emphasized. The calculated results establish confidence in the YAEC in-house computational methodology for general pressure vessel fluence analyses.

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

    NASA Astrophysics Data System (ADS)

    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.

  18. Nuclear technology aspects of ITER vessel-mounted diagnostics

    NASA Astrophysics Data System (ADS)

    Vayakis, George; Bertalot, Luciano; Encheva, Anna; Walker, Chris; Brichard, Benoît; Cheon, M. S.; Chitarin, G.; Hodgson, Eric; Ingesson, Christian; Ishikawa, M.; Kondoh, T.; Meister, Hans; Moreau, Philippe; Peruzzo, Simone; Pak, S.; Pérez-Pichel, Germán; Reichle, Roger; Testa, Duccio; Toussaint, Matthieu; Vermeeren, Ludo; Vershkov, Vladimir

    2011-10-01

    ITER has diagnostics with machine protection, basic and advanced control, and physics roles. Several are distributed on the inner and outer periphery of the vacuum vessel. They have reduced maintainability compared to diagnostics in ports. They also endure some of the highest nuclear and EM loads of any diagnostic for the longest time. They include: Inductive sensors for time-integrated and raw inductive measurements; Steady-state magnetic sensors to correct drifts of the inductive sensors; Bolometer cameras to provide electromagnetic radiation tomography; Microfission chambers and neutron activation stations to provide fusion power and fluence; MM-wave reflectometry to measure the plasma density profile and the plasma-wall distance and; Wiring to service magnetics, bolometry, and in-vessel instrumentation. This paper summarises the key technological issues these diagnostics arising from the nuclear environment, recent progress and outstanding R&D for each system.

  19. Simulating the Mineral Scale by High Pressure Thermal Vessel

    NASA Astrophysics Data System (ADS)

    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.

  20. IMPACT OF NUCLEAR MATERIAL DISSOLUTION ON VESSEL CORROSION

    SciTech Connect

    Mickalonis, J.; Dunn, K.; Clifton, B.

    2012-10-01

    Different nuclear materials require different processing conditions. In order to maximize the dissolver vessel lifetime, corrosion testing was conducted for a range of chemistries and temperature used in fuel dissolution. Compositional ranges of elements regularly in the dissolver were evaluated for corrosion of 304L, the material of construction. Corrosion rates of AISI Type 304 stainless steel coupons, both welded and non-welded coupons, were calculated from measured weight losses and post-test concentrations of soluble Fe, Cr and Ni.

  1. Numerical simulation of hydrogen diffusion in the pressure vessel wall of a WWER-440 reactor

    NASA Astrophysics Data System (ADS)

    Toribio, J.; Vergara, D.; Lorenzo, M.

    2017-07-01

    Materials forming the wall of a nuclear reactor pressure vessel (NRPV) can undergo in-service failure due to the presence of hydrogen, which enhances the fracture process known as hydrogen embrittlement (HE). A common way of avoiding this damage phenomenon is using a cladding material at the vessel wall side exposed to the hydrogenating source. This layer acts as a barrier for hydrogen diffusion and, hence, it protects the base material. In this paper, a numerical model of hydrogen diffusion assisted by stress and strain is used to analyse the hydrogen distribution, and hence the HE, in the pressure vessel wall of a real widely spread WWER-440 reactor considering two thickness for the cladding layer. Results show how the hydrogen accumulation is delayed as the thickness of the cladding layer increases, thus delaying the HE phenomenon affecting the structural integrity of the reactor.

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

    NASA Technical Reports Server (NTRS)

    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.

  3. Threaded insert for compact cryogenic-capable pressure vessels

    DOEpatents

    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.

  4. Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) Nickel-Hydrogen Battery Performance Under LEO Cycling Conditions

    NASA Technical Reports Server (NTRS)

    Miller, Thomas B.; Lewis, Harlan L.

    2004-01-01

    LEO life cycle testing of Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) nickel-hydrogen cell packs have been sponsored by the NASA Aerospace Flight Battery Program. The cell packs have cycled under both 35% and 60% depth-of- discharge and temperature conditions of -5 C and +lO C. The packs have been on test since as early as 1992 and have generated a substantial database. This report will provide insight into performance trends as a function of the specific cell configuration and manufacturer for eight separate nickel-hydrogen battery cell packs.

  5. Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) Nickel-Hydrogen Battery Performance Under LEO Cycling Conditions

    NASA Technical Reports Server (NTRS)

    Miller, Thomas B.; Lewis, Harlan L.

    2004-01-01

    LEO life cycle testing of Individual Pressure Vessel (PV) and Common Pressure Vessel (CPV) nickel-hydrogen cell packs have been sponsored by the NASA Aerospace Flight Battery Program. The cell packs have cycled under both 35% and 60% depth-of- discharge and temperature conditions of -5 C and +lO C. The packs have been on test since as early as 1992 and have generated a substantial database. This report will provide insight into performance trends as a function of the specific cell configuration and manufacturer for eight separate nickel-hydrogen battery cell packs.

  6. Low Temperature and High Pressure Evaluation of Insulated Pressure Vessels for Cryogenic Hydrogen Storage

    SciTech Connect

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

    2000-06-25

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

  8. Multiple cell common pressure vessel nickel hydrogen battery

    NASA Technical Reports Server (NTRS)

    Zagrodnik, Jeffrey P.; Jones, Kenneth R.

    1991-01-01

    A multiple cell common pressure vessel (CPV) nickel hydrogen battery was developed that offers significant weight, volume, cost, and interfacing advantages over the conventional individual pressure vessel (IPV) nickel hydrogen configuration that is currently used for aerospace applications. The baseline CPV design was successfully demonstrated though the testing of a 26 cell prototype, which completed over 7,000 44 percent depth of discharge LEO cycles. Two-cell boilerplate batteries have now exceeded 12,500 LEO cycles in ongoing laboratory tests. CPV batteries using both nominal 5 and 10 inch diameter vessels are currently available. The flexibility of the design allows these diameters to provide a broad capability for a variety of space applications.

  9. 46 CFR 50.30-15 - Class II pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS... Officer in Charge, Marine Inspection. The inspections described in this section are required, unless... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the marine...

  10. 46 CFR 50.30-15 - Class II pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS... Officer in Charge, Marine Inspection. The inspections described in this section are required, unless... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the marine...

  11. 46 CFR 50.30-15 - Class II pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS... Officer in Charge, Marine Inspection. The inspections described in this section are required, unless... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the marine...

  12. 46 CFR 50.30-15 - Class II pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS... Officer in Charge, Marine Inspection. The inspections described in this section are required, unless... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the marine...

  13. 46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 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...

  14. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 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...

  15. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 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...

  16. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 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...

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

    Code of Federal Regulations, 2010 CFR

    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 97.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 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...

  19. 46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 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...

  20. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 7 2014-10-01 2014-10-01 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...

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

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 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...

  2. Neutron-irradiated model alloys and pressure-vessel steels studied using positron spectroscopy

    NASA Astrophysics Data System (ADS)

    Cumblidge, Stephen Eric

    We have used positron-annihilation-lifetime spectroscopies to examine microstructural evolution of pressure vessel steels and model alloys that have systematically varied amounts of copper, nickel, and phosphorus during neutron irradiation and post-irradiation annealing. The objective of this work was to characterize the neutron-irradiation induced microstructural features that cause the embrittlement of nuclear reactor pressure-vessel steel. We used positron annihilation lifetime spectroscopy and Doppler-broadening spectroscopy to examine the model alloys and pressure-vessel steels before and after irradiation and after post-irradiation annealing. We followed the changes in the mechanical properties of the materials using Rockwell 15N hardness measurements. The results show that in both the model alloys and pressure-vessel steels neutron irradiation causes the formation of vacancy-type defect clusters and a fine distribution of copper- and nickel-enriched metallic precipitates. The vacancy clusters are small in size and were present in all samples, and disappear upon annealing at 450°C. The metallic precipitates are present only in the model alloy samples with either high Cu or a combination of medium Cu and high Ni, and they remain in the microstructure after annealing up to 550°C, starting to anneal possibly at 600°C. The neutron-irradiated pressure vessel steels behave similarly to the high Cu samples, indicating that neutron irradiation induced precipitation occurs in these alloys as well. This work provides independent evidence for the irradiation-induced metallic precipitates seen by other techniques, gives evidence for the exact nature of the matrix damage, and is significant to understanding the in-service degradation of pressure vessel materials.

  3. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology Program Series 4 and 5)

    SciTech Connect

    Berggren, R.G.; McGowan, J.J.; Menke, B.H.; Nanstad, R.K.; Thoms, K.R.

    1984-01-01

    Multiple testing is done at two laboratories of typical nuclear pressure vessel materials (both irradiated and unirradiated) and statistical analyses of the test results. Multiple tests are conducted at each of several test temperatures for each material, standard deviations are determined, and results from the two laboratories are compared. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (current practice welds). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds.

  4. Relating surveillance capsule measurements to pressure vessel damage

    SciTech Connect

    Carew, J.F.; Min, D.K.; Aronson, A.L.

    1980-01-01

    As part of the pressure vessel (PV) materials surveillance program, surveillance capsules including material specimens and neutron flux dosimeters are generally required to monitor changes in the fracture toughness properties of the reactor vessel materials. These capsules are withdrawn sequentially according to a predetermined schedule covering the service life of the vessel, and specimen material changes and dosimeter activation measured. The neutron fluence accumulated by the flux dosimeters is determined from the measured dosimeter activation and known reaction cross section (in practice, the /sup 54/Fe(n,p)/sup 54/Mn reaction.) The capsule fluence and material changes are then extrapolated to the pressure vessel using a fluence lead-factor determined from detailed multigroup neutron transport calculations. Typically, in this extrapolation changes in neutron spectrum are neglected. The purpose of this study is twofold; first, to determine the effect of including spectral changes in the extrapolation from capsule to vessel and second, to evaluate the effect of using the latest ENDF/B-V /sup 54/Fe(n,p)/sup 54/Mn cross sections in converting dosimeter activation to fluence.

  5. Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

    NASA Astrophysics Data System (ADS)

    Yi, Ping; Wang, Qingkang; Kong, Xianjing

    2017-01-01

    The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

  6. Lightweight pressure vessels and unitized regenerative fuel cells

    SciTech Connect

    Mitlitsky, F.; Myers, B.; Weisberg, A.H.

    1996-12-31

    High specific energy (>400 Wh/kg) energy storage systems have been designed using lightweight pressure vessels in conjunction with unitized regenerative fuel cells (URFCs). URFCs produce power and electrolytically regenerate their reactants using a single stack of reversible cells. Although a rechargeable energy storage system with such high specific energy has not yet been fabricated, we have made progress towards this goal. A primary fuel cell (FC) test rig with a single cell (0.05 ft{sup 2} active area) has been modified and operated reversibly as a URFC. This URFC uses bifunctional electrodes (oxidation and reduction electrodes reverse roles when switching from charge to discharge, as with a rechargeable battery) and cathode feed electrolysis (water is fed from the oxygen side of the cell). Lightweight pressure vessels with state-of-the-art performance factors (burst pressure * internal volume/tank weight = Pb V/W) have been designed and fabricated. These vessels provide a lightweight means of storing reactant gases required for fuel cells (FCs) or URFCs. The vessels use lightweight bladder liners that act as inflatable mandrels for composite overwrap and provide the permeation barrier for gas storage. The bladders are fabricated using materials that are compatible with humidified gases which may be created by the electrolysis of water and are compatible with elevated temperatures that occur during fast fills.

  7. Reliability Considerations for Composite Overwrapped Pressure Vessels on Spacecraft

    NASA Technical Reports Server (NTRS)

    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.

  8. Material Issues in Space Shuttle Composite Overwrapped Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Sutter, James K.; Jensen, Brian J.; Gates, Thomas S.; Morgan, Roger J.; Thesken, John C.; Phoenix, S. Leigh

    2006-01-01

    Composite Overwrapped Pressure Vessels (COPV) store gases used in four subsystems for NASA's Space Shuttle Fleet. While there are 24 COPV on each Orbiter ranging in size from 19-40", stress rupture failure of a pressurized Orbiter COPV on the ground or in flight is a catastrophic hazard and would likely lead to significant damage/loss of vehicle and/or life and is categorized as a Crit 1 failure. These vessels were manufactured during the late 1970's and into the early 1980's using Titanium liners, Kevlar 49 fiber, epoxy matrix resin, and polyurethane coating. The COPVs are pressurized periodically to 3-5ksi and therefore experience significant strain in the composite overwrap. Similar composite vessels were developed in a variety of DOE Programs (primarily at Lawrence Livermore National Laboratories or LLNL), as well as for NASA Space Shuttle Fleet Leader COPV program. The NASA Engineering Safety Center (NESC) formed an Independent Technical Assessment (ITA) team whose primary focus was to investigate whether or not enough composite life remained in the Shuttle COPV in order to provide a strategic rationale for continued COPV use aboard the Space Shuttle Fleet with the existing 25-year-old vessels. Several material science issues were examined and will be discussed in this presentation including morphological changes to Kevlar 49 fiber under stress, manufacturing changes in Kevlar 49 and their effect on morphology and tensile strength, epoxy resin strain, composite creep, degradation of polyurethane coatings, and Titanium yield characteristics.

  9. Improved Attachment in a Hybrid Inflatable Pressure Vessel

    NASA Technical Reports Server (NTRS)

    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.

  10. Fabrication of toroidal composite pressure vessels. Final report

    SciTech Connect

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

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

    NASA Technical Reports Server (NTRS)

    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.

  13. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 4 2014-10-01 2014-10-01 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...

  14. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 4 2013-10-01 2013-10-01 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...

  15. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 4 2011-10-01 2011-10-01 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...

  16. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 4 2012-10-01 2012-10-01 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...

  17. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 197.462 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE OCCUPATIONAL SAFETY... to the satisfaction of the Officer in Charge, Marine Inspection. (c) The following tests shall be... shall be submitted to the Officer in Charge, Marine Inspection. (d) Unless otherwise noted, pressure...

  18. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 197.462 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE OCCUPATIONAL SAFETY... to the satisfaction of the Officer in Charge, Marine Inspection. (c) The following tests shall be... shall be submitted to the Officer in Charge, Marine Inspection. (d) Unless otherwise noted, pressure...

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

    NASA Technical Reports Server (NTRS)

    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.

  20. Residual Stress Measurements of Explosively Clad Cylindrical Pressure Vessels

    SciTech Connect

    Taylor, Douglas J; Watkins, Thomas R; Hubbard, Camden R; Hill, M. R.; Meith, W. A.

    2012-01-01

    Tantalum refractory liners were explosively clad into cylindrical pressure vessels, some of which had been previously autofrettaged. Using explosive cladding, the refractory liner formed a metallurgical bond with the steel of the pressure vessel at a cost of induced strain. Two techniques were employed to determine the residual stress state of the clad steel cylinders: neutron diffraction and mechanical slitting. Neutron diffraction is typically nondestructive; however, due to attenuation along the beam path, the cylinders had to be sectioned into rings that were nominally 25 mm thick. Slitting is a destructive method, requiring the sectioning of the cylindrical samples. Both techniques provided triaxial stress data and useful information on the effects of explosive cladding. The stress profiles in the hoop and radial directions were similar for an autofrettaged, nonclad vessel and a clad, nonautofrettaged vessel. The stress profiles in the axial direction appeared to be different. Further, the data suggested that residual stresses from the autofrettage and explosive cladding processes were not additive, in part due to evidence of reverse yielding. The residual stress data are presented, compared and discussed.

  1. Niobium Application, Metallurgy and Global Trends in Pressure Vessel Steels

    NASA Astrophysics Data System (ADS)

    Jansto, Steven G.

    Niobium-containing high strength steel materials have been developed for a variety of pressure vessel applications. Through the application of these Nb-bearing steels in demanding applications, the designer and end user experience improved toughness at low temperature, excellent fatigue resistance and fracture toughness and excellent weldability. These enhancements provide structural engineers the opportunity to further improve the pressure vessel design and performance. The Nb-microalloy alloy designs also result in reduced operational production cost at the steel operation, thereby embracing the value-added attribute Nb provides to both the producer and the end user throughout the supply chain. For example, through the adoption of these Nb-containing structural materials, several design-manufacturing companies are considering improved designs which offer improved manufacturability, lower overall cost and better life cycle performance.

  2. Glass Fiber Reinforced Metal Pressure Vessel Design Guide

    NASA Technical Reports Server (NTRS)

    Landes, R. E.

    1972-01-01

    The Engineering Guide presents curves and general equations for safelife design of lightweight glass fiber reinforced (GFR) metal pressure vessels operating under anticipated Space Shuttle service conditions. The high composite vessel weight efficiency is shown to be relatively insensitive to shape, providing increased flexibility to designers establishing spacecraft configurations. Spheres, oblate speroids, and cylinders constructed of GFR Inconel X-750, 2219-T62 aluminum, and cryoformed 301 stainless steel are covered; design parameters and performance efficiencies for each configuration are compared at ambient and cryogenic temperature for an operating pressure range of 690 to 2760 N/sq cm (1000 to 4000 psi). Design variables are presented as a function of metal shell operating to sizing (proof) stress ratios for use with fracture mechanics data generated under a separate task of this program.

  3. Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs)

    NASA Technical Reports Server (NTRS)

    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.

  4. Lightweight pressure vessels and unitized regenerative fuel cells

    SciTech Connect

    Mitlitsky, F.; Myers, B.; Weisberg, A.H.

    1996-09-06

    Energy storage systems have been designed using lightweight pressure vessels with unitized regenerative fuel cells (URFCs). The vessels provide a means of storing reactant gases required for URFCs; they use lightweight bladder liners that act as inflatable mandrels for composite overwrap and provide a permeation barrier. URFC systems have been designed for zero emission vehicles (ZEVs); they are cost competitive with primary FC powered vehicles that operate on H/air with capacitors or batteries for power peaking and regenerative braking. URFCs are capable of regenerative braking via electrolysis and power peaking using low volume/low pressure accumulated oxygen for supercharging the power stack. URFC ZEVs can be safely and rapidly (<5 min.) refueled using home electrolysis units. Reversible operation of cell membrane catalyst is feasible without significant degradation. Such systems would have a rechargeable specific energy > 400 Wh/kg.

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

    SciTech Connect

    Theofanous, T.G.; Syri, S.

    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.

  6. Prediction of radiation induced hardening of reactor pressure vessel steels using artificial neural networks

    NASA Astrophysics Data System (ADS)

    Castin, N.; Malerba, L.; Chaouadi, R.

    2011-01-01

    In this paper, we use an artificial neural network approach to obtain predictions of neutron irradiation induced hardening, more precisely of the change in the yield stress, for reactor pressure vessel steels of pressurized water nuclear reactors. Different training algorithms are proposed and compared, with the goal of identifying the best procedure to follow depending on the needs of the user. The numerical importance of some input variables is also studied. Very accurate numerical regressions are obtained, by taking only four input variables into account: neutron fluence, irradiation temperature, and chemical composition (Cu and Ni content). Accurate extrapolations in term of neutron fluence are obtained.

  7. Thermally activated deformation of irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    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.

  8. Design of pressure vessel cascades for electromagnetic launchers

    SciTech Connect

    Fahrenthold, E.P. )

    1989-08-01

    The relatively recent development of very high-energy density pulsed power supplies has motivated a renewed interest in the structural design of electromagnetic launchers. Cascade design electromagnetic launcher pressure vessels offer convenient maintenance access to high wear rate components of the structure while satisfying an unusual combination of electromagnetic, strength, and preloading constraints imposed on the system designer. This analysis for design of such structures focuses on the accurate characterization of fluid-structure interaction under dynamic asymmetric loading.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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 are discussed as well as the research and development activities conducted at ORNL in support of PCPV development.

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

    NASA Technical Reports Server (NTRS)

    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.

  12. Temperature and pressure influence on explosion pressures of closed vessel propane-air deflagrations.

    PubMed

    Razus, Domnina; Brinzea, Venera; Mitu, Maria; Oancea, Dumitru

    2010-02-15

    An experimental study on pressure evolution during closed vessel explosions of propane-air mixtures was performed, for systems with various initial concentrations and pressures ([C(3)H(8)]=2.50-6.20 vol.%, p(0)=0.3-1.2 bar). The explosion pressures and explosion times were measured in a spherical vessel (Phi=10 cm), at various initial temperatures (T(0)=298-423 K) and in a cylindrical vessel (Phi=10 cm; h=15 cm), at ambient initial temperature. The experimental values of explosion pressures are examined against literature values and compared to adiabatic explosion pressures, computed by assuming chemical equilibrium within the flame front. The influence of initial pressure, initial temperature and fuel concentration on explosion pressures and explosion times are discussed. At constant temperature and fuel/oxygen ratio, the explosion pressures are linear functions of total initial pressure, as reported for other fuel-air mixtures. At constant initial pressure and composition, both the measured and calculated (adiabatic) explosion pressures are linear functions of reciprocal value of initial temperature. Such correlations are extremely useful for predicting the explosion pressures of flammable mixtures at elevated temperatures and/or pressures, when direct measurements are not available.

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

    NASA Technical Reports Server (NTRS)

    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.

  14. Terahertz NDE of Stressed Composite Overwrapped Pressure Vessels - Initial Testing

    NASA Technical Reports Server (NTRS)

    Madaras, Eric I.; Seebo, Jeffrey P.; Anatasi, Robert F.

    2009-01-01

    Terahertz radiation nondestructive evaluation was applied to a set of Kevlar composite overwrapped pressure vessel bottles that had undergone a series of thermal and pressure tests to simulate stress rupture effects. The bottles in these nondestructive evaluation tests were bottles that had not ruptured but had survived various times at the elevated load and temperature levels. Some of the bottles showed evidence of minor composite failures. The terahertz radiation did detect visible surface flaws, but did not detect any internal chemical or material degradation of the thin overwraps.

  15. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... they have been designed and constructed to meet the standards of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section VIII, Rules for Construction of Unfired Pressure Vessels...

  16. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... they have been designed and constructed to meet the standards of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section VIII, Rules for Construction of Unfired Pressure Vessels...

  17. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... they have been designed and constructed to meet the standards of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section VIII, Rules for Construction of Unfired Pressure Vessels...

  18. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... they have been designed and constructed to meet the standards of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section VIII, Rules for Construction of Unfired Pressure Vessels...

  19. Dual shell pressure balanced reactor vessel. Final project report

    SciTech Connect

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy`s Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R&D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993).

  20. Three-dimensional neutron flux calculations for the VVER pressure vessel

    SciTech Connect

    Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

    1995-08-01

    The neutron flux values at the sites important for the pressure vessels of the VVER-1000 and VVER-440 reactors have been calculated by the three-dimensional TORT code and the synthesis method approximation. The synthesis method is widely used now for neutron fluence routine calculations in metal embrittlement surveillance. The three-dimensional neutron flux evaluation by the synthesis method is based on the two-dimensional and one-dimensional solutions of the transport equation. The comparison of the results obtained by both methods confirms the good consistency within 3% for integral neutron flux with energy >0.5 MeV, used for metal damage estimation, according to Russian reactor standards. Further investigations on the calculation validity will be based on comparisons with measurements of the threshold detector activities, monitored in the air shell behind the reactor pressure vessels of the Kozloduy nuclear power plant.

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

    NASA Astrophysics Data System (ADS)

    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.

  2. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Code of Federal Regulations, 2010 CFR

    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...

  3. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 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...

  4. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 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...

  5. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 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)...

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

    Code of Federal Regulations, 2010 CFR

    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 78.33-1 - Repairs of boiler and pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 3 2012-10-01 2012-10-01 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...

  8. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 7 2011-10-01 2011-10-01 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...

  9. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Code of Federal Regulations, 2010 CFR

    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...

  10. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 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)...

  11. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 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...

  12. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 7 2013-10-01 2013-10-01 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)...

  13. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 3 2013-10-01 2013-10-01 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...

  14. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 3 2014-10-01 2014-10-01 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...

  15. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 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...

  16. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 7 2012-10-01 2012-10-01 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...

  17. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 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...

  18. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 3 2011-10-01 2011-10-01 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...

  19. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 5 2012-10-01 2012-10-01 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...

  20. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 5 2011-10-01 2011-10-01 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...

  1. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 5 2013-10-01 2013-10-01 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...

  2. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 5 2014-10-01 2014-10-01 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...

  3. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54.01-10 Section 54.01-10 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g...

  4. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54.01-10 Section 54.01-10 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g...

  5. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54.01-10 Section 54.01-10 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g...

  6. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54.01-10 Section 54.01-10 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g...

  7. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 56.13015 Section 56.13015 Mineral Resources MINE SAFETY AND HEALTH...

  8. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 56.13015 Section 56.13015 Mineral Resources MINE SAFETY AND HEALTH...

  9. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 56.13015 Section 56.13015 Mineral Resources MINE SAFETY AND HEALTH...

  10. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2010 CFR

    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...

  11. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 56.13015 Section 56.13015 Mineral Resources MINE SAFETY AND HEALTH...

  12. PRESSURIZATION OF CONTAINMENT VESSELS FROM PLUTONIUM OXIDE CONTENTS

    SciTech Connect

    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.

  13. Jam proof closure assembly for lidded pressure vessels

    DOEpatents

    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.

  14. The rapid depressurization of initially saturated fluids from pressure vessels

    NASA Astrophysics Data System (ADS)

    Fishman, R. E.

    1980-03-01

    Nonequilibrium effects during rapid depressurizations include large initial pressure fluctuations, a separated flow pattern consisting of a superheated vapor region above a two phase boiling froth, a significant temperature gradient in the vapor zone and entrained liquid droplets in the effluent stream. The physical basis for the nonequilibrium behavior was examined. A model which predicts these phenomena as well as the pressure time relationship for depressurizing vessels containing boiling fluids is described. The submodels include: (1) a Second Law 'tracer' which analyzes the thermodynamic process followed by an open, depressurizing system; (2) a correlation predicting the wall superheat required to support nucleation during rapid depressurizations; (3) a correlation establishing the relationship between the exit mass flow rate and the void fraction in the boiling region; (4) an analysis which predicts droplet entrainment by vapor bubbles rising in a boiling froth; and (5) an analysis which predicts the initial transient behavior. The results of the model were compared with Freon-12 depressurization experiments utilizing small pressure vessels.

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

    NASA Technical Reports Server (NTRS)

    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.

  16. Composite Overwrapped Pressure Vessels (COPV) Materials Aging Issues

    NASA Technical Reports Server (NTRS)

    2010-01-01

    This slide presentation reviews some of the issues concerning the aging of the materials in a Composite Overwrapped Pressure Vessels (COPV). The basic composition of the COPV is a Boss, a composite overwrap, and a metallic liner. The lifetime of a COPV is affected by the age of the overwrap, the cyclic fatigue of the metallic liner, and stress rupture life, a sudden and catastrophic failure of the overwrap while holding at a stress level below the ultimate strength for an extended time. There is information about the coupon tests that were performed, and a test on a flight COPV.

  17. Fabrication Flaws in Reactor Pressure Vessel Repair Welds

    SciTech Connect

    Schuster, George J.; Doctor, Steven R.

    2007-12-01

    This paper describes the fabrication flaw distribution and characterization in the repair weld metal of reactor pressure vessels. This work indicates that the large flaws occur in these repairs. These results show that repair flaws are complex in composition and sometimes include cracks on the repair ends. Parametric analysis using an exponential fit is performed on the data. A description of repair flaw morphology is provided. Fabrication flaws in repairs are characterized using high sensitivity nondestructive ultrasonic testing, validation by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing.

  18. Evaluation of the reactor pressure vessel steels by positron annihilation

    NASA Astrophysics Data System (ADS)

    Slugeň, V.; Hein, H.; Sojak, S.; Simeg Veterníková, J.; Petriska, M.; Sabelová, V.; Pavúk, M.; Hinca, R.; Stacho, M.

    2013-11-01

    This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades. The second generation of Russian RPV steels seems to be fully comparable with German steels and their quality allows prolongation of NPP operating lifetime over projected 40 years. The embrittlement of CrMoV steels is relatively low due to effect of higher temperature which implies partial in situ annealing of primary microstructural point defects and therefore delays the degradation processes caused by neutron irradiation.

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

    NASA Astrophysics Data System (ADS)

    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.

  20. Strain measurements using FBG on composite over wrap pressure vessels (COPV) in stress rupture test

    NASA Astrophysics Data System (ADS)

    Grant, Joseph; Banks, Curtis

    2007-04-01

    Thirty six Fiber Optic Braggs Grating sensors were used during an ambient temperature hydrostatic pressurization testing of a Space Transportation System (STS) 40-inch Kevlar Composite Over-wrapped Pressure Vessel (COPV). The 40-inch vessel was of the same design and approximate age as the STS Main Propulsion System (MPS) and Orbiter Maneuvering System (OMS) vessels. The sensors were surfaces mounted to on the vessel to measure strain during a stress rupture event. The Bragg signals were linear with the applied pressure. The results indicated that the vessel was under an uneven force distribution at various locations on the vessel.

  1. Influence of crack depth on the fracture toughness of reactor pressure vessel steel

    SciTech Connect

    Theiss, T.J.; Bryson, J.W.

    1991-01-01

    The Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. Recently, it has been shown that, in notched beam testing, shallow cracks tend to exhibit an elevated toughness as a result of a loss of constraint at the crack tip. The loss of constraint takes place when interaction occurs between the elastic-plastic crack-tip stress field and the specimen surface nearest the crack tip. An increased shallow-crack fracture toughness is of interest to the nuclear industry because probabilistic fracture-mechanics evaluations show that shallow flaws play a dominant role in the probability of vessel failure during postulated pressurized-thermal-shock (PTS) events. Tests have been performed on beam specimens loaded in 3-point bending using unirradiated reactor pressure vessel material (A533 B). Testing has been conducted using specimens with a constant beam depth (W = 94 mm) and within the lower transition region of the toughness curve for A533 B. Test results indicate a significantly higher fracture toughness associated with the shallow flaw specimens compared to the fracture toughness determined using deep-crack (a/W = 0.5) specimens. Test data also show little influence of thickness on the fracture toughness for the current test temperature ({minus}60{degree}C). 21 refs., 5 figs., 3 tabs.

  2. Making a Metal-Lined Composite-Overwrapped Pressure Vessel

    NASA Technical Reports Server (NTRS)

    DeLay, Tom

    2005-01-01

    process has been devised for the fabrication of a pressure vessel that comprises a composite-material (matrix/fiber) shell with a metal liner on its inner surface. The use of the composite material makes it possible for the tank to be strong enough to withstand the anticipated operating pressure and yet weigh less than does an equivalent all-metal tank. The metal liner is used as a barrier against permeation: In the absence of such a barrier, the pressurized gas in the tank could leak by diffusing through the composite-material shell. The figure depicts workpieces at four key stages in the process, which consists of the following steps: 1. A mandrel that defines the size and shape of the pressure vessel is made by either molding or machining a piece of tooling wax. 2. Silver paint is applied to the surface of the mandrel to make it electrically conductive. 3. The ends of the mandrel are fitted with metal bosses. 4. The mandrel is put into a plating bath, wherein the metal liner is electrodeposited. Depending on the applications, the liner metal could be copper, nickel, gold, or an alloy. Typical liner thicknesses range from 1 to 10 mils (0.025 to 0.25 mm). 5. The wax is melted from within, leaving the thin metal liner. 6. A hollow shaft that includes holes and fittings through which the liner can be pressurized is sealed to both ends of the liner. The liner is pressurized to stiffen (and hence stabilize) it for the next step. 7. The pressurized liner is placed in a filament-winding machine, which is then operated to cover the liner with multiple layers of an uncured graphite-fiber/epoxy-matrix or other suitable composite material. 8. The composite-overwrapped liner is cured in an oven. 9. The pressure is relieved and the shaft is removed. The tank is then ready for use. The process as described above accommodates variations: a) The mandrel could be made of a wax that melts at a higher temperature and not removed until the tank is cured in the oven. b) The tank need

  3. Stress distribution in continuously heterogeneous thick laminated pressure vessels

    SciTech Connect

    Verijenko, V.E.; Adali, S.; Tabakov, P.Y.

    1995-11-01

    Stress analysis of multilayered pressure vessels possessing cylindrical anisotropy and under internal, external and interlaminar pressure is given. The special case when the axis of anisotropy coincides with the axis of symmetry Oz and the stresses do not vary long the generator is investigated. In this case there exists a plane of elastic symmetry normal to this axis at every point of the cylinder so that each layer may be considered s orthotropic. However, elastic properties can vary through the thickness of a layer. Exact elasticity solutions are obtained for both open-ended and closed-ended cylinders using a stress function approach. The method of solution allows the forces on the layer interfaces to be taken into account with relative ease. Numerical results are presented for thick cylinders with isotropic and orthotropic layers, and stress distributions across the thickness are given.

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

    NASA Technical Reports Server (NTRS)

    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.

  5. Magnetic Barkhausen noise and magneto acoustic emission in pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Neyra Astudillo, Miriam Rocío; López Pumarega, María Isabel; Núñez, Nicolás Marcelo; Pochettino, Alberto; Ruzzante, José

    2017-03-01

    Magnetic Barkhausen Noise (MBN) and Magneto Acoustic Emission (MAE) were studied in A508 Class II forged steel used for pressure vessels in nuclear power stations. The magnetic experimental determinations were completed with a macro graphic study of sulfides and the texture analysis of the material. The analysis of these results allows us to determine connections between the magnetic anisotropy, texture and microstructure of the material. Results clearly suggest that the plastic flow direction is different from the forging direction indicated by the material supplier

  6. Magnetic non-destructive evaluation of hardening of cold rolled reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Wang, Xuejiao; Qiang, Wenjiang; Shu, Guogang

    2017-08-01

    Non-destructive test (NDT) of reactor pressure vessel (RPV) steel is urgently required due to the life extension program of nuclear power plant. Here magnetic NDT of cold rolled RPV steel is studied. The strength, hardness and coercivity increase with the increasing deformation, and a good linear correlation between the increment of coercivity, hardness and yield strength is found, which may be helpful to develop magnetic NDT of degradation of RPV steel. It is also found that besides dislocation density, the distribution of dislocations may affect coercivity as well.

  7. Mechanical properties and examination of cracking in TMI-2 pressure vessel lower head material

    SciTech Connect

    Diercks, D.R.; Neimark, L.A.

    1993-09-01

    Mechanical tests have been conducted on material from 15 samples removed from the lower head of the Three Mile Island unit 2 nuclear reactor pressure vessel. Measured properties include tensile properties and hardness profiles at room temperature, tensile and creep properties at temperatures of 600 to 1200{degrees}C, and Charpy V-notch impact properties at {minus}20 to +300{degrees}C. These data, which were used in the subsequent analyses of the margin-to-failure of the lower head during the accident, are presented here. In addition, the results of metallographic and scanning electron microscope examinations of cladding cracking in three of the lower head samples are discussed.

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

    SciTech Connect

    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.

  9. Finite element analysis of filament-wound composite pressure vessel under internal pressure

    NASA Astrophysics Data System (ADS)

    Sulaiman, S.; Borazjani, S.; Tang, S. H.

    2013-12-01

    In this study, finite element analysis (FEA) of composite overwrapped pressure vessel (COPV), using commercial software ABAQUS 6.12 was performed. The study deals with the simulation of aluminum pressure vessel overwrapping by Carbon/Epoxy fiber reinforced polymer (CFRP). Finite element method (FEM) was utilized to investigate the effects of winding angle on filament-wound pressure vessel. Burst pressure, maximum shell displacement and the optimum winding angle of the composite vessel under pure internal pressure were determined. The Laminae were oriented asymmetrically for [00,00]s, [150,-150]s, [300,-300]s, [450,-450]s, [550,-550]s, [600,-600]s, [750,-750]s, [900,-900]s orientations. An exact elastic solution along with the Tsai-Wu, Tsai-Hill and maximum stress failure criteria were employed for analyzing data. Investigations exposed that the optimum winding angle happens at 550 winding angle. Results were compared with the experimental ones and there was a good agreement between them.

  10. Flux effect analysis in WWER-440 reactor pressure vessel steels

    NASA Astrophysics Data System (ADS)

    Kryukov, A.; Blagoeva, D.; Debarberis, L.

    2013-11-01

    The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013-0.036 wt%) and copper (0.08-0.20 wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ˜4 × 1012 сm-2 s-1 and low ˜6 × 1011 сm-2 s-1 fluxes (E > 0.5 MeV) at a temperature of 270 °С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25-50 °C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13 wt%).

  11. Stresses in reactor pressure vessel nozzles -- Calculations and experiments

    SciTech Connect

    Brumovsky, M.; Polachova, H.

    1995-11-01

    Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber`s, Hardrath-Ohman`s as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared.

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

    SciTech Connect

    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.

  13. A Survey of Pressure Vessel Code Compliance for Superconducting RF Cryomodules

    SciTech Connect

    Peterson, Thomas; Klebaner, Arkadiy; Nicol, Tom; Theilacker, Jay; Hayano, Hitoshi; Kako, Eiji; Nakai, Hirotaka; Yamamoto, Akira; Jensch, Kay; Matheisen, Axel; Mammosser, John; /Jefferson Lab

    2011-06-07

    Superconducting radio frequency (SRF) cavities made from niobium and cooled with liquid helium are becoming key components of many particle accelerators. The helium vessels surrounding the RF cavities, portions of the niobium cavities themselves, and also possibly the vacuum vessels containing these assemblies, generally fall under the scope of local and national pressure vessel codes. In the U.S., Department of Energy rules require national laboratories to follow national consensus pressure vessel standards or to show ''a level of safety greater than or equal to'' that of the applicable standard. Thus, while used for its superconducting properties, niobium ends up being treated as a low-temperature pressure vessel material. Niobium material is not a code listed material and therefore requires the designer to understand the mechanical properties for material used in each pressure vessel fabrication; compliance with pressure vessel codes therefore becomes a problem. This report summarizes the approaches that various institutions have taken in order to bring superconducting RF cryomodules into compliance with pressure vessel codes. In Japan, Germany, and the U.S., institutions building superconducting RF cavities integrated in helium vessels or procuring them from vendors have had to deal with pressure vessel requirements being applied to SRF vessels, including the niobium and niobium-titanium components of the vessels. While niobium is not an approved pressure vessel material, data from tests of material samples provide information to set allowable stresses. By means of procedures which include adherence to code welding procedures, maintaining material and fabrication records, and detailed analyses of peak stresses in the vessels, or treatment of the vacuum vessel as the pressure boundary, research laboratories around the world have found methods to demonstrate and document a level of safety equivalent to the applicable pressure vessel codes.

  14. Prediction of Composite Pressure Vessel Failure Location using Fiber Bragg Grating Sensors

    NASA Technical Reports Server (NTRS)

    Kreger, Steven T.; Taylor, F. Tad; Ortyl, Nicholas E.; Grant, Joseph

    2006-01-01

    Ten composite pressure vessels were instrumented with fiber Bragg grating sensors in order to assess the strain levels of the vessel under various loading conditions. This paper and presentation will discuss the testing methodology, the test results, compare the testing results to the analytical model, and present a possible methodology for predicting the failure location and strain level of composite pressure vessels.

  15. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... vessels, built after the effective date of this regulation, shall be marked and reported indicating that... pressure of the vessels. (b) Portable, unfired pressure vessels, not built to the code requirements of paragraph (a) of this section, and built prior to the effective date of this regulation, shall be...

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

    NASA Astrophysics Data System (ADS)

    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.

  17. Thin-metal lined PRD 49-III composite vessels. [evaluation of pressure vessels for burst strength and fatigue performance

    NASA Technical Reports Server (NTRS)

    Hoggatt, J. T.

    1974-01-01

    Filament wound pressure vessels of various configurations were evaluated for burst strength and fatigue performance. The dimensions and characteristics of the vessels are described. The types of tests conducted are explained. It was determined that all vessels leaked in a relatively few cycles (20 to 60 cycles) with failure occurring in all cases in the metallic liner. The thin liner would de-bond from the composite and buckling took place during depressurization. No composite failures or indications of impeding composite failures were obtained in the metal-lined vessels.

  18. Regulatory Activities Related to Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles

    SciTech Connect

    Hiser, Allen L. Jr.

    2002-07-01

    The recent discoveries of cracked and leaking Alloy 600 vessel head penetration (VHP) nozzles, including control rod drive mechanism (CRDM) and thermocouple nozzles, at four pressurized water reactors (PWRs) have raised concerns about the structural integrity of VHP nozzles throughout the PWR industry. Nozzle cracking at Oconee Nuclear Station Unit 1 in November 2000 and Arkansas Nuclear One Unit 1 in February 2001 was limited to axial cracking, an occurrence deemed to be of limited safety concern in the NRC staff's generic safety evaluation on the cracking of VHP nozzles dated November 19, 1993. However, the discovery of circumferential cracking at Oconee Nuclear Station Unit 3 in February 2001 and Oconee Nuclear Station Unit 2 in April 2001 particularly the large circumferential cracking identified in two CRDM nozzles at ONS3 has raised concerns about the potential safety implications and prevalence of cracking in VHP nozzles in PWRs. In response to the circumferential cracking identified at the Oconee units, the NRC issued Bulletin 2001-01, 'Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles', on August 3, 2001. This bulletin requests information from licensees related to the structural integrity of the reactor pressure VHP nozzles for their facilities, including the extent of VHP nozzle leakage and cracking that has been found to date, the inspections and repairs that have been undertaken to satisfy applicable regulatory requirements, future plans to inspect VHP nozzles, and a description of how future inspection plans will ensure compliance with applicable regulatory requirements. This paper summarizes the staff's review and assessment of licensee responses to NRC Bulletin 2001-01. (author)

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

    NASA Technical Reports Server (NTRS)

    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).

  20. Structural Health Monitoring of Composite Wound Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Grant, Joseph; Kaul, Raj; Taylor, Scott; Jackson, Kurt; Myers, George; Sharma, A.

    2002-01-01

    The increasing use of advanced composite materials in the wide range of applications including Space Structures is a great impetus to the development of smart materials. Incorporating these FBG sensors for monitoring the integrity of structures during their life cycle will provide valuable information about viability of the usage of such material. The use of these sensors by surface bonding or embedding in this composite will measure internal strain and temperature, and hence the integrity of the assembled engineering structures. This paper focuses on such a structure, called a composite wound pressure vessel. This vessel was fabricated from the composite material: TRH50 (a Mitsubishi carbon fiber with a 710-ksi tensile strength and a 37 Msi modulus) impregnated with an epoxy resin from NEWPORT composites (WDE-3D-1). This epoxy resin in water dispersed system without any solvents and it cures in the 240-310 degrees F range. This is a toughened resin system specifically designed for pressure applications. These materials are a natural fit for fiber sensors since the polyimide outer buffer coating of fiber can be integrated into the polymer matrix of the composite material with negligible residual stress. The tank was wound with two helical patterns and 4 hoop wraps. The order of winding is: two hoops, two helical and two hoops. The wall thickness of the composite should be about 80 mil or less. The tank should burst near 3,000 psi or less. We can measure the actual wall thickness by ultrasonic or we can burst the tank and measure the pieces. Figure 1 shows a cylinder fabricated out of carbon-epoxy composite material. The strain in different directions is measured with a surface bonded fiber Bragg gratings and with embedded fiber Bragg gratings as the cylinder is pressurized to burst pressures. Figure 2 shows the strain as a function of pressure of carbon-epoxy cylinder as it is pressurized with water. Strain is measured in different directions by multiple gratings

  1. Structural Health Monitoring of Composite Wound Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Grant, Joseph; Kaul, Raj; Taylor, Scott; Jackson, Kurt; Myers, George; Sharma, A.

    2002-01-01

    The increasing use of advanced composite materials in the wide range of applications including Space Structures is a great impetus to the development of smart materials. Incorporating these FBG sensors for monitoring the integrity of structures during their life cycle will provide valuable information about viability of the usage of such material. The use of these sensors by surface bonding or embedding in this composite will measure internal strain and temperature, and hence the integrity of the assembled engineering structures. This paper focuses on such a structure, called a composite wound pressure vessel. This vessel was fabricated from the composite material: TRH50 (a Mitsubishi carbon fiber with a 710-ksi tensile strength and a 37 Msi modulus) impregnated with an epoxy resin from NEWPORT composites (WDE-3D-1). This epoxy resin in water dispersed system without any solvents and it cures in the 240-310 degrees F range. This is a toughened resin system specifically designed for pressure applications. These materials are a natural fit for fiber sensors since the polyimide outer buffer coating of fiber can be integrated into the polymer matrix of the composite material with negligible residual stress. The tank was wound with two helical patterns and 4 hoop wraps. The order of winding is: two hoops, two helical and two hoops. The wall thickness of the composite should be about 80 mil or less. The tank should burst near 3,000 psi or less. We can measure the actual wall thickness by ultrasonic or we can burst the tank and measure the pieces. Figure 1 shows a cylinder fabricated out of carbon-epoxy composite material. The strain in different directions is measured with a surface bonded fiber Bragg gratings and with embedded fiber Bragg gratings as the cylinder is pressurized to burst pressures. Figure 2 shows the strain as a function of pressure of carbon-epoxy cylinder as it is pressurized with water. Strain is measured in different directions by multiple gratings

  2. Simply actuated closure for a pressure vessel - Design for use to trap deep-sea animals

    NASA Technical Reports Server (NTRS)

    Yayanos, A. A.

    1977-01-01

    A pressure vessel is described that can be closed by a single translational motion within 1 sec. The vessel is a key component of a trap for small marine animals and operates automatically on the sea floor. As the vessel descends to the sea floor, it is subjected both internally and externally to the high pressures of the deep sea. The mechanism for closing the pressure vessel on the sea floor is activated by the timed release of the ballast which was used to sink the trap. As it rises to the sea surface, the internal pressure of the vessel remains near the value present on the sea floor. The pressure vessel has been used in simulated ocean deployments and in the deep ocean (9500 m) with a 75%-85% retention of the deep-sea pressure. Nearly 100% retention of pressure can be achieved by using an accumulator filled with a gas.

  3. Burst prediction by acoustic emission in filament-wound pressure vessels

    NASA Technical Reports Server (NTRS)

    Gorman, Michael R.

    1990-01-01

    Acoustic emission in 51-cm diameter graphite/epoxy pressure vessels was monitored during pressurization (hydrotesting). Several vessels were subjected to impact by a blunt impactor, but only after the vessels had been proofed; that is, pressurized to 80 percent of nominal burst pressure as determined from control (unimpacted) vessels. AE activity was then monitored throughout a series of successively higher pressure cycles ranging from 10 to 60 percent of ultimate. Each cycle included a ramp up to pressure followed by a 4-min hold period and then pressure unload. The event rate was high, and especially modified AE analyzers had to be used to acquire the data. This paper presents the AE event count versus pressure history of these tests and demonstrates the ability of the AE technique to monitor the growth of damage and to estimate the effect on ultimate strength. The number of events that occurred during pressure holds proved to be a reasonable estimator of vessel performance.

  4. Burst prediction by acoustic emission in filament-wound pressure vessels

    NASA Technical Reports Server (NTRS)

    Gorman, Michael R.

    1990-01-01

    Acoustic emission in 51-cm diameter graphite/epoxy pressure vessels was monitored during pressurization (hydrotesting). Several vessels were subjected to impact by a blunt impactor, but only after the vessels had been proofed; that is, pressurized to 80 percent of nominal burst pressure as determined from control (unimpacted) vessels. AE activity was then monitored throughout a series of successively higher pressure cycles ranging from 10 to 60 percent of ultimate. Each cycle included a ramp up to pressure followed by a 4-min hold period and then pressure unload. The event rate was high, and especially modified AE analyzers had to be used to acquire the data. This paper presents the AE event count versus pressure history of these tests and demonstrates the ability of the AE technique to monitor the growth of damage and to estimate the effect on ultimate strength. The number of events that occurred during pressure holds proved to be a reasonable estimator of vessel performance.

  5. Design and Analysis of Boiler Pressure Vessels based on IBR codes

    NASA Astrophysics Data System (ADS)

    Balakrishnan, B.; Kanimozhi, B.

    2017-05-01

    Pressure vessels components are widely used in the thermal and nuclear power plants for generating steam using the philosophy of heat transfer. In Thermal power plant, Coal is burnt inside the boiler furnace for generating the heat. The amount of heat produced through the combustion of pulverized coal is used in changing the phase transfer (i.e. Water into Super-Heated Steam) in the Pressure Parts Component. Pressure vessels are designed as per the Standards and Codes of the country, where the boiler is to be installed. One of the Standards followed in designing Pressure Parts is ASME (American Society of Mechanical Engineers). The mandatory requirements of ASME code must be satisfied by the manufacturer. In our project case, A Shell/pipe which has been manufactured using ASME code has an issue during the drilling of hole. The Actual Size of the drilled holes must be, as per the drawing, but due to error, the size has been differentiate from approved design calculation (i.e. the diameter size has been exceeded). In order to rectify this error, we have included an additional reinforcement pad to the drilled and modified the design of header in accordance with the code requirements.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  8. Techniques for Embedding Instrumentation in Pressure Vessel Test Articles

    NASA Technical Reports Server (NTRS)

    Cornelius, Michael

    2006-01-01

    Many interesting structural and thermal events occur in materials that are housed within a surrounding pressure vessel. In order to measure the environment during these events and explore their causes instrumentation must be installed on or in the material. Transducers can be selected that are small enough to be embedded within the test material but these instruments must interface with an external system in order to apply excitation voltages and output the desired data. The methods for installing the instrumentation and creating an interface are complicated when the material is located in a case or housing containing high pressures and hot gases. Installation techniques for overcoming some of these difficulties were developed while testing a series of small-scale solid propellant and hybrid rocket motors at Marshall Space Flight Center. These techniques have potential applications in other test articles where data are acquired from materials that require containment due to the severe environment encountered during the test process. This severe environment could include high pressure, hot gases, or ionized atmospheres. The development of these techniques, problems encountered, and the lessons learned from the ongoing testing process are summarized.

  9. Techniques for embedding instrumentation in pressure vessel test articles

    NASA Astrophysics Data System (ADS)

    Cornelius, Michael

    2006-05-01

    Many interesting structural and thermal events occur in materials that are housed within a surrounding pressure vessel. In order to measure the environment during these events and explore their causes instrumentation must be installed on or in the material. Transducers can be selected that are small enough to be embedded within the test material but these instruments must interface with an external system in order to apply excitation voltages and output the desired data. The methods for installing the instrumentation and creating an interface are complicated when the material is located in a case or housing containing high pressures and hot gases. Installation techniques for overcoming some of these difficulties were developed while testing a series of small-scale solid propellant and hybrid rocket motors at Marshall Space Flight Center. These techniques have potential applications in other test articles where data are acquired from materials that require containment due to the severe environment encountered during the test process. This severe environment could include high pressure, hot gases, or ionized atmospheres. The development of these techniques, problems encountered, and the lessons learned from the ongoing testing process are summarized.

  10. Recent advances in lightweight, filament-wound composite pressure vessel technology

    NASA Technical Reports Server (NTRS)

    Lark, R. F.

    1977-01-01

    A review of recent advances is presented for lightweight, high performance composite pressure vessel technology that covers the areas of design concepts, fabrication procedures, applications, and performance of vessels subjected to single cycle burst and cyclic fatigue loading. Filament wound fiber/epoxy composite vessels were made from S glass, graphite, and Kevlar 49 fibers and were equipped with both structural and nonstructural liners. Pressure vessels structural efficiencies were attained which represented weight savings, using different liners, of 40 to 60 percent over all titanium pressure vessels. Significant findings in each area are summarized.

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

    SciTech Connect

    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 which

  12. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    NASA Technical Reports Server (NTRS)

    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

  13. ACS Algorithm in Discrete Ordinates for Pressure Vessel Dosimetry

    NASA Astrophysics Data System (ADS)

    Walters, William; Haghighat, Alireza

    2016-02-01

    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.

  14. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    NASA Technical Reports Server (NTRS)

    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

  15. Macrosegregation and Microstructural Evolution in a Pressure-Vessel Steel

    NASA Astrophysics Data System (ADS)

    Pickering, E. J.; Bhadeshia, H. K. D. H.

    2014-06-01

    This work assesses the consequences of macrosegregation on microstructural evolution during solid-state transformations in a continuously cooled pressure-vessel steel (SA508 Grade 3). Stark spatial variations in microstructure are observed following a simulated quench from the austenitization temperature, which are found to deliver significant variations in hardness. Partial-transformation experiments are used to show the development of microstructure in segregated material. Evidence is presented which indicates the bulk microstructure is not one of upper bainite, as it has been described in the past, but one comprised of Widmanstätten ferrite and pockets of lower bainite. Segregation is observed on three different length scales, and the origins of each type are proposed. Suggestions are put forward for how the segregation might be minimized, and its detrimental effects suppressed by heat treatments.

  16. Young's modulus anisotropy in reactor pressure vessel cladding

    NASA Astrophysics Data System (ADS)

    Vandermeulen, W.; Mertens, M.; Scibetta, M.

    2012-02-01

    In a previous study it was shown that the anisotropy of Young's modulus in the stainless steel cladding of a reactor pressure vessel could be attributed to the solidification texture of the cladding. Further it was found that annealing the samples to remove the delta phase caused a modulus change but only in some directions. Since the texture was only estimated from X-ray diffraction patterns the moduli, calculated for some principal directions, differed considerably from the measured ones. In the present study, executed on a practically identical cladding, the texture was determined by actual texture measurements. It was found to be close to a fibre texture with <0 0 1> perpendicular to the cladding plane and the values calculated from it agreed much better with the experimental ones. The annealing effect found in the previous study was shown to be due to surface recrystallization induced by milling damage.

  17. Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

    SciTech Connect

    Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

    2014-02-18

    Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

  18. Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

    NASA Astrophysics Data System (ADS)

    Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

    2014-02-01

    Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

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

    SciTech Connect

    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.

  20. Marine transportation and burial of the Shippingport reactor pressure vessel/neutron shield tank package

    SciTech Connect

    Coughlin, P.J.

    1989-01-01

    The Shippingport Station Decommissioning Project (SSDP) is a US Department of Energy (DOE) project for dismantling the Shippingport atomic power station. One of the more significant and challenging technical aspects of the project, which is being managed for DOE by General Electric-Nuclear Energy, is the marine transport of the reactor pressure vessel (RPV) and its associated neutron shield tank (NST) to the government-owned Hanford Reservation near Richland, Washington. Planning of the transport activity, barge transportation operations, and Hanford transportation operations, are discussed. This work will be the first use of barge transportation in the United States of a radioactive RPV package from a decommissioned land-based nuclear power plant. This extensive transportation operation has been accomplished in a timely, safe, and cost-effective manner.

  1. Polymer-lined filament-wound pressure vessels for nitrogen containment

    NASA Technical Reports Server (NTRS)

    Hamstad, M. A.; Chiao, T. T.; Jessop, E. S.

    1974-01-01

    A program has been started to develop fatigue-resistant polymeric liners for a filament-wound pressure vessel to contain nitrogen gas at room temperature. First, nitrogen permeation of butyl rubber sheet coated with Saran and Parylene C was studied in flat specimens. Then four 10-cm-diam cylindrical pressure vessels were prepared with chlorobutyl rubber liners coated with the same materials. These vessels were valved off after nitrogen gas pressurization to approximately 65% (approximately 11.7 MPa or 1700 psig) of their expected failure pressure. One vessel leaked. The other three vessels showed an average pressure loss of less than 1% per month. These pressure vessels have an average performance factor of approximately 370 kPa-cu m/kg (1,500,000 in.) based on composite mass.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

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

    SciTech Connect

    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

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

    NASA Technical Reports Server (NTRS)

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

    2007-01-01

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

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

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-27

    ...] Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying Liquefied... announces the availability of CG-ENG Policy Letter 04-12, ``Alternative Pressure Relief Valve Settings on... Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII with respect...

  7. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance...

  8. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance...

  9. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance...

  10. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 30 Mineral Resources 1 2013-07-01 2013-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

  11. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

  12. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 30 Mineral Resources 1 2012-07-01 2012-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

  13. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance...

  14. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance...

  15. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 30 Mineral Resources 1 2014-07-01 2014-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

  16. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 30 Mineral Resources 1 2011-07-01 2011-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

  17. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2010 CFR

    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...

  18. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure... 30 Mineral Resources 1 2011-07-01 2011-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 57.13015 Section 57.13015 Mineral Resources MINE SAFETY AND HEALTH...

  19. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure... 30 Mineral Resources 1 2012-07-01 2012-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 57.13015 Section 57.13015 Mineral Resources MINE SAFETY AND HEALTH...

  20. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure... 30 Mineral Resources 1 2013-07-01 2013-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 57.13015 Section 57.13015 Mineral Resources MINE SAFETY AND HEALTH...

  1. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure... 30 Mineral Resources 1 2014-07-01 2014-07-01 false Inspection of compressed-air receivers and other unfired pressure vessels. 57.13015 Section 57.13015 Mineral Resources MINE SAFETY AND HEALTH...

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

    SciTech Connect

    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.

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

    SciTech Connect

    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.

  4. Development and Demonstration of Insulated Pressure Vessels for Vehicular Hydrogen Storage

    SciTech Connect

    Berry, G D; Aceves, S M

    2004-02-26

    This paper describes the development of an alternative technology for vehicular storage of hydrogen. Insulated pressure vessels are cryogenic-capable pressure vessels that can accept cryogenic liquid fuel, cryogenic compressed gas or compressed gas at ambient temperature. Insulated pressure vessels offer advantages over alternative hydrogen storage technologies. Insulated pressure vessels are more compact and less expensive than compressed hydrogen vessels. They have lower evaporative losses and lower energy requirement for fuel liquefaction than liquid hydrogen tanks, and they are lighter than hydrides. The work described in this paper is directed at verifying that insulated pressure vessels can be used safely for vehicular hydrogen storage. The paper describes multiple tests and analyses that have been conducted to evaluate the safety of insulated pressure vessels. Insulated pressure vessels have been subjected to multiple DOT, ISO and SAE certification tests, and the vessels have always been successful in meeting the passing criteria for the different tests. A draft procedure for insulated pressure vessel certification has been generated to assist in a future commercialization of this technology. Ongoing work includes the demonstration of this technology in a vehicle.

  5. Structural considerations in design of lightweight glass-fiber composite pressure vessels

    NASA Technical Reports Server (NTRS)

    Faddoul, J. R.

    1973-01-01

    The development of structurally efficient, metal-lined, glass-fiber composite pressure vessels. Both the current state-of-the-art and current problems are discussed along with fracture mechanics considerations for the metal liner. The design concepts used for metal-lined, glass-fiber, composite pressure vessels are described and the structural characteristics of the composite designs are compared with each other and with homogeneous metal pressure vessels. Specific design techniques and available design data are identified. Results of a current program to evaluate flaw growth and fracture characteristics of the metal liners are reviewed and the impact of these results on composite pressure vessel designs is discussed.

  6. The Development of Radiation Embrittlement Models for U. S. Power Reactor Pressure Vessel Steels

    SciTech Connect

    Wang, Jy-An John; Rao, Nageswara S; Konduri, Savanthi

    2007-01-01

    A new approach of utilizing information fusion technique is developed to predict the radiation embrittlement of reactor pressure vessel steels. The Charpy transition temperature shift data contained in the Power Reactor Embrittlement Database is used in this study. Six parameters {Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature {are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

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

    SciTech Connect

    Qian, Xudong; Dodds, Robert; Yin, Shengjun; Bass, Bennett Richard

    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

  8. Filament wound pressure vessels with load sharing liners for Space Shuttle Orbiter applications

    NASA Technical Reports Server (NTRS)

    Ecord, G. M.

    1976-01-01

    It is recognized that the use of overwrapped pressure vessels with load sharing liners may provide significant weight savings for high pressure gas containment in Space Shuttle Orbiter systems. The technology readiness to produce Kevlar wound vessels with load sharing liners of titanium 6Al-4V, Inconel 718 or cryoformed 301 steel has been demonstrated. It has been estimated that about 400 lbs can be saved in the Orbiter by using overwrapped vessels with load sharing liners instead of monolithic metal designs. Total weight of the composite vessels would be about 1350 lbs as opposed to about 1750 lbs for all-metal vessels.

  9. Predictive Reactor Pressure Vessel Steel Irradiation Embrittlement Models: Issues and Opportunities

    SciTech Connect

    Odette, George Robert; Nanstad, Randy K

    2009-01-01

    Nuclear plant life extension to 80 years will require accurate predictions of neutron irradiation-induced increases in the ductile-brittle transition temperature ( T) of reactor pressure vessel (RPV) steels at high fluence conditions that are far outside the existing database. Remarkable progress in mechanistic understanding of irradiation embrittlement has led to physically motivated T correlation models that provide excellent statistical fi ts to the existing surveillance database. However, an important challenge is developing advanced embrittlement models for low fl ux-high fl uence conditions pertinent to extended life. These new models must also provide better treatment of key variables and variable combinations and account for possible delayed formation of late blooming phases in low copper steels. Other issues include uncertainties in the compositions of actual vessel steels, methods to predict T attenuation away from the reactor core, verifi cation of the master curve method to directly measure the fracture toughness with small specimens and predicting T for vessel annealing remediation and re-irradiation cycles.

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

    SciTech Connect

    Spencer, Benjamin; Backman, Marie; Williams, Paul; Hoffman, William; Alfonsi, Andrea; Dickson, Terry; Bass, B. Richard; Klasky, Hilda

    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.

  11. A Neural Network/Acoustic Emission Analysis of Impact Damaged Graphite/Epoxy Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Walker, James L.; Hill, Erik v. K.; Workman, Gary L.; Russell, Samuel S.

    1995-01-01

    Acoustic emission (AE) signal analysis has been used to measure the effects of impact damage on burst pressure in 5.75 inch diameter, inert propellant filled, filament wound pressure vessels. The AE data were collected from fifteen graphite/epoxy pressure vessels featuring five damage states and three resin systems. A burst pressure prediction model was developed by correlating the AE amplitude (frequency) distribution, generated during the first pressure ramp to 800 psig (approximately 25% of the average expected burst pressure for an undamaged vessel) to known burst pressures using a four layered back propagation neural network. The neural network, trained on three vessels from each resin system, was able to predict burst pressures with a worst case error of 5.7% for the entire fifteen bottle set.

  12. Pressure-vessel-damage fluence reduction by low-leakage fuel management. [PWR

    SciTech Connect

    Cokinos, D.; Aronson, A.L.; Carew, J.F.; Kohut, P.; Todosow, M.; Lois, L.

    1983-01-01

    As a result of neutron-induced radiation damage to the pressure vessel and of an increased concern that in a PWR transient the pressure vessel may be subjected to pressurized thermal shock (PTS), detailed analyses have been undertaken to determine the levels of neutron fluence accumulation at the pressure vessels of selected PWR's. In addition, various methods intended to limit vessel damage by reducing the vessel fluence have been investigated. This paper presents results of the fluence analysis and the evaluation of the low-leakage fuel management fluence reduction method. The calculations were performed with DOT-3.5 in an octant of the core/shield/vessel configuration using a 120 x 43 (r, theta) mesh structure.

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

    NASA Astrophysics Data System (ADS)

    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

  14. Pressurized water nuclear reactor system with hot leg vortex mitigator

    DOEpatents

    Lau, Louis K. S.

    1990-01-01

    A pressurized water nuclear reactor system includes a vortex mitigator in the form of a cylindrical conduit between the hot leg conduit and a first section of residual heat removal conduit, which conduit leads to a pump and a second section of residual heat removal conduit leading back to the reactor pressure vessel. The cylindrical conduit is of such a size that where the hot leg has an inner diameter D.sub.1, the first section has an inner diameter D.sub.2, and the cylindrical conduit or step nozzle has a length L and an inner diameter of D.sub.3 ; D.sub.3 /D.sub.1 is at least 0.55, D.sub.2 is at least 1.9, and L/D.sub.3 is at least 1.44, whereby cavitation of the pump by a vortex formed in the hot leg is prevented.

  15. Pressure vessel fluence monitoring at NPP with VVER: Routine technique and new approaches

    SciTech Connect

    Borodkin, G.I.; Kovalevich, O.M.; Lomakin, S.S.; Sycheva, N.V.

    1994-12-31

    For Reactor Pressure Vessel (RPV) neutron fluence monitoring at a nuclear power plant (NPP) with VVER-type reactor the Russian Nuclear Regulatory Body (GOSATOMNADZOR) has recently recommended use of the ex-vessel cavity dosimetry combined with neutron transport calculations. By using activation dosimeters with long-lived reaction products on the base of {sup 54}Fe, {sup 58}Ni, {sup 46}Ti, {sup 60}Ni, {sup 63}Cu, {sup 55}Mn and {sup 59}Co the routine experiments are now carried out. The VVER cavity dosimetry methodology has been developed on the basis of experimental results obtained at Rovno NPP Unit 1 (Ro-1), Unit 2 (Ro-2), Unit 3 (Ro-3), Novovoronezh NPP Unit 5 (NV-5), Yuzhno-Ukrainsk NPP Unit 1 (YuU-1) with VVER-440 and VVER-1000. The set of experiments has been performed during the various reactor fuel cycles including the low leakage core. More accurate evaluation of integral neutron spectrum in cavity is possible by using short-time irradiations of the dosimeters with short-lived reaction products. Now performing cavity experiments also include {sup 237}Np, {sup 238}U and {sup 93}Nb detectors. The track detectors, similar SSTR, may be used for cavity dosimetry, and now such set dosimeters are available for installation in cavities of some VVERs. Also the on-line fluence monitoring technique is now developed. A summary of experimental and calculational results is presented.

  16. Photofission Analysis for Fissile Dosimeters Dedicated to Reactor Pressure Vessel Surveillance

    NASA Astrophysics Data System (ADS)

    Bourganel, Stéphane; Faucher, Margaux; Thiollay, Nicolas

    2016-02-01

    Fissile dosimeters are commonly used in reactor pressure vessel surveillance programs. In this paper, the photofission contribution is analyzed for in-vessel 237Np and 238U fissile dosimeters in French PWR. The aim is to reassess this contribution using recent tools (the TRIPOLI-4 Monte Carlo code) and latest nuclear data (JEFF3.1.1 and ENDF/B-VII nuclear libraries). To be as exhaustive as possible, this study is carried out for different configurations of fissile dosimeters, irradiated inside different kinds of PWR: 900 MWe, 1300 MWe, and 1450 MWe. Calculation of photofission rate in dosimeters does not present a major problem using the TRIPOLI-4® Monte Carlo code and the coupled neutron-photon simulation mode. However, preliminary studies were necessary to identify the origin of photons responsible of photofissions in dosimeters in relation to the photofission threshold reaction (around 5 MeV). It appears that the main contribution of high enough energy photons generating photofissions is the neutron inelastic scattering in stainless steel reactor structures. By contrast, 137Cs activity calculation is not an easy task since photofission yield data are known with high uncertainty.

  17. A nonintrusive nuclear monitor for measuring liquid contents in sealed vessels

    NASA Technical Reports Server (NTRS)

    Singh, J. J.; Mall, G. H.

    1984-01-01

    A nonintrusive nuclear technique for monitoring fluid contents in sealed vessels, regardless of the fluid distribution inside the vessels is described. The technique is applicable to all-g environments. It is based on the differences in Cesium-137 gamma ray attenuation coefficients in air and the test liquids.

  18. Dual shell reactor vessel: A pressure-balanced system for high pressure and temperature reactions

    SciTech Connect

    Robertus, R.J.; Fassbender, A.G.; Deverman, G.S.

    1995-03-01

    The main purpose of this work was to demonstrate the Dual Shell Pressure Balanced Vessel (DSPBV) as a safe and economical reactor for the hydrothermal water oxidation of hazardous wastes. Experimental tests proved that the pressure balancing piston and the leak detection concept designed for this project will work. The DSPBV was sized to process 10 gal/hr of hazardous waste at up to 399{degree}C (750{degree}F) and 5000 psia (34.5 MPa) with a residence time of 10 min. The first prototype reactor is a certified ASME pressure vessel. It was purchased by Innotek Corporation (licensee) and shipped to Pacific Northwest Laboratory for testing. Supporting equipment and instrumentation were, to a large extent, transported here from Battelle Columbus Division. A special air feed system and liquid pump were purchased to complete the package. The entire integrated demonstration system was assembled at PNL. During the activities conducted for this report, the leak detector design was tested on bench top equipment. Response to low levels of water in oil was considered adequate to ensure safety of the pressure vessel. Shakedown tests with water only were completed to prove the system could operate at 350{degree}C at pressures up to 3300 psia. Two demonstration tests with industrial waste streams were conducted, which showed that the DSPBV could be used for hydrothermal oxidation. In the first test with a metal plating waste, chemical oxygen demand, total organic carbon, and cyanide concentrations were reduced over 90%. In the second test with a munitions waste, the organics were reduced over 90% using H{sub 2}O{sub 2} as the oxidant.

  19. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    SciTech Connect

    Manahan, M.P. Sr.

    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.

  20. Impact damage and burst of filament-wound CFRP composite pressure vessel

    SciTech Connect

    Matemilola, S.A.; Stronge, W.J.

    1996-12-31

    Quasi-static and impact tests were conducted on filament-wound carbon fiber composite pressure vessels to study factors that affect burst pressure. Observed damage include fiber microbuckling, matrix cracking, and delamination. For vessels that were not pressurized during test, both the matrix cracking and fiber breakage were restricted to the outer layer, whereas in the case of an internally pressurized vessel struck by a wedge nose shaped impactor these cracks extended into the second layer. Fiber microbuckling of the outer surface layer near the impact point was the main factor that degraded the burst pressure of the vessels. This type of damage was visually detectable on the surface. For an unpressurized vessel it appeared as a pair of cracks radiating from the periphery of contact region. On the other hand, for a pressurized vessel circumferential microbuckling developed within the contact region. The burst pressure for a damaged vessel decreased as the ratio of axial length of the buckled fibers l, to vessel thickness h, increased, up to a ratio {ell}/h {approx} 3, beyond which the burst pressure became constant. Strain measurements near the region of loading showed that fiber microbuckling occurred, the failure strain value at a strain rate of 104 s{sup {minus}1} was about six times the microbuckling strain for quasi-static loading.

  1. Experimental Investigation of Composite Pressure Vessel Performance and Joint Stiffness for Pyramid and Inverted Pyramid Joints

    NASA Technical Reports Server (NTRS)

    Verhage, Joseph M.; Bower, Mark V.; Gilbert, Paul A. (Technical Monitor)

    2001-01-01

    The focus of this study is on the suitability in the application of classical laminate theory analysis tools for filament wound pressure vessels with adhesive laminated joints in particular: pressure vessel wall performance, joint stiffness and failure prediction. Two 18-inch diameter 12-ply filament wound pressure vessels were fabricated. One vessel was fabricated with a 24-ply pyramid laminated adhesive double strap butt joint. The second vessel was fabricated with the same number of plies in an inverted pyramid joint. Results from hydrostatic tests are presented. Experimental results were used as input to the computer programs GENLAM and Laminate, and the output compared to test. By using the axial stress resultant, the classical laminate theory results show a correlation within 1% to the experimental results in predicting the pressure vessel wall pressure performance. The prediction of joint stiffness for the two adhesive joints in the axial direction is within 1% of the experimental results. The calculated hoop direction joint stress resultant is 25% less than the measured resultant for both joint configurations. A correction factor is derived and used in the joint analysis. The correction factor is derived from the hoop stress resultant from the tank wall performance investigation. The vessel with the pyramid joint is determined to have failed in the joint area at a hydrostatic pressure 33% value below predicted failure. The vessel with the inverted pyramid joint failed in the wall acreage at a hydrostatic pressure within 10% of the actual failure pressure.

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

    NASA Astrophysics Data System (ADS)

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  4. Effect of Macrosegregation on the Microstructure and Mechanical Properties of a Pressure-Vessel Steel

    NASA Astrophysics Data System (ADS)

    Yan, Guanghua; Han, Lizhan; Li, Chuanwei; Luo, Xiaomeng; Gu, Jianfeng

    2017-07-01

    Macrosegregation refers to the chemical segregation, which occurs quite commonly in the large forgings such as nuclear reactor pressure vessel. This work assesses the effect of macrosegregation and homogenization treatment on the mechanical properties of a pressure-vessel steel (SA508 Gr.3). It was found that the primary reason for the inhomogeneity of the microstructure was the segregation of Mn, Mo, and Ni. Martensite, and coarse upper bainite with M-A (martensite-austenite) islands have been obtained, respectively, in the positive and negative segregation zone during a simulated quenching process. During tempering, the carbon-rich M-A islands decomposed into a mixture of ferrite and numerous carbides which deteriorated the toughness of the material. The segregation has been substantially minimized by a homogenizing treatment. The results indicate that the material homogenized has a higher impact toughness than the material with segregation, due to the reduction in M-A island in the negative segregation zone. It can be concluded that the microstructure and mechanical properties have been improved remarkably by means of homogenization treatment.

  5. Determination of the Critical Buckling Pressure of Blood Vessels Using the Energy Approach

    PubMed Central

    Han, Hai-Chao

    2011-01-01

    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. PMID:21116846

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

    PubMed

    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.

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

    NASA Astrophysics Data System (ADS)

    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.

  8. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOEpatents

    Challberg, R.C.; Gou, P.F.; Chu, C.L.; Oliver, R.P.

    1999-07-27

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block. 6 figs.

  9. Assemblies and methods for mitigating effects of reactor pressure vessel expansion

    DOEpatents

    Challberg, Roy C.; Gou, Perng-Fei; Chu, Cherk Lam; Oliver, Robert P.

    1999-01-01

    Support assemblies for allowing RPV radial expansion while simultaneously limiting horizontal, vertical, and azimuthal movement of the RPV within a nuclear reactor are described. In one embodiment, the support assembly includes a support block and a guide block. The support block includes a first portion and a second portion, and the first portion is rigidly coupled to the RPV adjacent the first portion. The guide block is rigidly coupled to a reactor pressure vessel support structure and includes a channel sized to receive the second portion of the support block. The second portion of the support block is positioned in the guide block channel to movably couple the guide block to the support block.

  10. Underclad cracking of pressure vessel steels for light-water reactors

    SciTech Connect

    Lopez, H.F.

    1987-06-01

    Although fracture mechanics analyses have shown that underclad cracks have no detrimental effect on the integrity of thick walled pressure vessels (40 year service), in order to avoid unexpected failures the US Nuclear Regulatory Commission has issued Regulatory Guide 1.43 which sets limits on the extent of fissures permitted and describes acceptable means of controlling the weld cladding processes. Cavitation and intergranular fissuring in SA508-2 and 22NiMoCr37 steels can occur in the presence or absence of intergranular particles. The observations of intergranular fissuring and cavitation in those HAZ free from overlapping effects are attributed to grain boundary segregation. Other probable void nucleation sites are the grain boundary-lath interface intersections which facilitate the formation of grain boundary discontinuities.

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

    NASA Astrophysics Data System (ADS)

    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.

  12. REACTOR PRESSURE VESSEL ISSUES FOR THE LIGHT-WATER REACTOR SUSTAINABILITY PROGRAM

    SciTech Connect

    Nanstad, Randy K; Odette, George Robert

    2010-01-01

    The Light Water Reactor Sustainability Program Plan is a collaborative program between the U.S. Department of Energy and the private sector directed at extending the life of the present generation of nuclear power plants to enable operation to at least 80 years. The reactor pressure vessel (RPV) is one of the primary components requiring significant research to enable such long-term operation. There are significant issues that need to be addressed to reduce the uncertainties in regulatory application, such as, 1) high neutron fluence/long irradiation times, and flux effects, 2) material variability, 3) high-nickel materials, 4)specimen size effects and the fracture toughness master curve, etc. The first issue is the highest priority to obtain the data and mechanistic understanding to enable accurate, reliable embrittlement predictions at high fluences. This paper discusses the major issues associated with long-time operation of existing RPVs and the LWRSP plans to address those issues.

  13. Some Observations on Damage Tolerance Analyses in Pressure Vessels

    NASA Technical Reports Server (NTRS)

    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

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

    SciTech Connect

    Spencer, Benjamin; Hoffman, William; Sen, Sonat; Rabiti, Cristian; Dickson, Terry; Bass, Richard

    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

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

    SciTech Connect

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    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

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

    SciTech Connect

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    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.

  17. Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

    SciTech Connect

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

    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.

  18. Low Pressure Nuclear Thermal Rocket (LPNTR) concept

    NASA Technical Reports Server (NTRS)

    Ramsthaler, J. H.

    1991-01-01

    A background and a description of the low pressure nuclear thermal system are presented. Performance, mission analysis, development, critical issues, and some conclusions are discussed. The following subject areas are covered: LPNTR's inherent advantages in critical NTR requirement; reactor trade studies; reference LPNTR; internal configuration and flow of preliminary LPNTR; particle bed fuel assembly; preliminary LPNTR neutronic study results; multiple LPNTR engine concept; tank and engine configuration for mission analysis; LPNTR reliability potential; LPNTR development program; and LPNTR program costs.

  19. A dual output pressure, high reliability, long storage life gas delivery vessel assembly

    NASA Astrophysics Data System (ADS)

    Maya, Isaac; McKee, Joe; Rajpurkar, Rajiv

    1993-02-01

    A Gas Vessel Assembly has been developed that delivers purified, very low moisture content gas at two different output pressures. High pressure gas is delivered at up to 6,700 psi, and low pressure gas regulated to 130 psi is also delivered via a second outlet over a wide range of flow rates. The device is extremely lightweight (less than 1 lb) and compact, affords maximum mechanical integrity, high reliability (0.9999 at 95 percent confidence level), and offers extremely long storage life. Specialized design and fabrication techniques are employed that guarantee gas purity and negligible leakage for more than 20 years, in widely varying conditions of storage temperature, humidity, altitude, and vibration environments. The technology offers unique advantages in fast, high pressure discharge applications. For example, when combined with a cryostat, cryogenic temperatures can be achieved such as those used in missile seeker technology. The technology has many additional applications such as: emergency power sources for safety devices such as those needed in nuclear power plants, refineries, collision cushioning devices, superconductor cooling devices, emergency egress systems, miniature mechanical devices that employ gas bearings, and other areas where long storage, extremely high reliability and/or high energy density sources are required.

  20. A dual output pressure, high reliability, long storage life gas delivery vessel assembly

    NASA Technical Reports Server (NTRS)

    Maya, Isaac; Mckee, Joe; Rajpurkar, Rajiv

    1993-01-01

    A Gas Vessel Assembly has been developed that delivers purified, very low moisture content gas at two different output pressures. High pressure gas is delivered at up to 6,700 psi, and low pressure gas regulated to 130 psi is also delivered via a second outlet over a wide range of flow rates. The device is extremely lightweight (less than 1 lb) and compact, affords maximum mechanical integrity, high reliability (0.9999 at 95 percent confidence level), and offers extremely long storage life. Specialized design and fabrication techniques are employed that guarantee gas purity and negligible leakage for more than 20 years, in widely varying conditions of storage temperature, humidity, altitude, and vibration environments. The technology offers unique advantages in fast, high pressure discharge applications. For example, when combined with a cryostat, cryogenic temperatures can be achieved such as those used in missile seeker technology. The technology has many additional applications such as: emergency power sources for safety devices such as those needed in nuclear power plants, refineries, collision cushioning devices, superconductor cooling devices, emergency egress systems, miniature mechanical devices that employ gas bearings, and other areas where long storage, extremely high reliability and/or high energy density sources are required.

  1. Photoacoustic sample vessel and method of elevated pressure operation

    DOEpatents

    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.

  2. Non-invasive method and apparatus for measuring pressure within a pliable vessel

    NASA Technical Reports Server (NTRS)

    Shimizu, M. (Inventor)

    1983-01-01

    A non-invasive method and apparatus is disclosed for measuring pressure within a pliable vessel such as a blood vessel. The blood vessel is clamped by means of a clamping structure having a first portion housing a pressure sensor and a second portion extending over the remote side of the blood vessel for pressing the blood vessel into engagement with the pressure sensing device. The pressure sensing device includes a flat deflectable diaphragm portion arranged to engage a portion of the blood vessel flattened against the diaphragm by means of the clamp structure. In one embodiment, the clamp structure includes first and second semicylindrical members held together by retaining rings. In a second embodiment the clamp structure is of one piece construction having a solid semicylindrical portion and a hollow semicylindrical portion with a longitudinal slot in the follow semicylindrical portion through which a slip the blood vessel. In a third embodiment, an elastic strap is employed for clamping the blood vessel against the pressure sensing device.

  3. Structural Integrity of Gas-Filled Composite Overwrapped Pressure Vessels Subjected to Orbital Debris Impact

    NASA Astrophysics Data System (ADS)

    Telichev, Igor; Cherniaev, Aleksandr

    Gas-filled pressure vessels are extensively used in spacecraft onboard systems. During operation on the orbit they exposed to the space debris environment. Due to high energies they contain, pressure vessels have been recognized as the most critical spacecraft components requiring protection from orbital debris impact. Major type of pressurized containers currently used in spacecraft onboard systems is composite overwrapped pressure vessels (COPVs) manufactured by filament winding. In the present work we analyze the structural integrity of vessels of this kind in case of orbital debris impact at velocities ranging from 2 to 10 km/s. Influence of such parameters as projectile energy, shielding standoff, internal pressure and filament winding pattern on COPVs structural integrity has been investigated by means of numerical and physical experiments.

  4. Evaluation of Progressive Failure Analysis and Modeling of Impact Damage in Composite Pressure Vessels

    NASA Technical Reports Server (NTRS)

    Sanchez, Christopher M.

    2011-01-01

    NASA White Sands Test Facility (WSTF) is leading an evaluation effort in advanced destructive and nondestructive testing of composite pressure vessels and structures. WSTF is using progressive finite element analysis methods for test design and for confirmation of composite pressure vessel performance. Using composite finite element analysis models and failure theories tested in the World-Wide Failure Exercise, WSTF is able to estimate the static strength of composite pressure vessels. Additionally, test and evaluation on composites that have been impact damaged is in progress so that models can be developed to estimate damage tolerance and the degradation in static strength.

  5. Low background stainless steel for the pressure vessel in the PandaX-II dark matter experiment

    NASA Astrophysics Data System (ADS)

    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.

  6. On-site implementation of characterization and sizing techniques for outer-wall defects in reactor pressure vessels

    SciTech Connect

    Lasserre, F.; Chapuis, N.

    1994-12-31

    Pressurized reactor vessels in France have been examined from the inside with ultrasonic focused transducers since the very first inspection. The developments carried out to solve the problem of oversizing in the case of defects located near the outer surface in the welds or in the wall thickness and presented in the framework of the 10th and 11th conference of NDE in the nuclear and pressure vessels industries, now have applications through SPARTACUS software work. Indications detected during, the systematic inspection of welds and shells, corresponding to outer wall defects, trigger a digital acquisition of data, the scanning being limited to the area of interest. This acquisition is now followed by analysis through the new system CIVAMIS, which includes the main imaging tools of SPARTACUS, but which has been specifically developed to be implemented on site, for outer wall defects. Characteristics of CIVAMIS in relation with the initial structure of SPARTACUS are discussed on actual results.

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

    SciTech Connect

    Marek, M.; Rataj, J.; Vandlik, S.

    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)

  8. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    SciTech Connect

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.; Wells, P.; Stan, T.; Almirall, N.; Odette, G. R.; Ecker, L. E.

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitates that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  9. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    DOE PAGES

    Sprouster, D. J.; Sinsheimer, J.; Dooryhee, E.; ...

    2015-10-21

    Here, massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that increases with dose (fluence or service time), as manifested by an increasing temperature transition from ductile-to-brittle fracture. Moreover, extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with 60 to 80 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize a new class of highly embrittling nm-scale Mn-Ni-Si precipitatesmore » that develop in the irradiated steels at high fluence. Furthermore, these precipitates can lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementarity techniques has, for the very first time, successfully characterized the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.« less

  10. Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

    SciTech Connect

    Lu, S.C.; Sommer, S.C.; Johnson, G.L. ); Lambert, H.E. )

    1990-10-01

    This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns.

  11. Recent advances in lightweight, filament-wound composite pressure vessel technology

    NASA Technical Reports Server (NTRS)

    Lark, R. F.

    1977-01-01

    A review of recent advances is presented for lightweight, high-performance composite pressure vessel technology that covers the areas of design concepts, fabrication procedures, applications, and performance of vessels subjected to single-cycle burst and cyclic fatigue loading. Filament-wound fiber/epoxy composite vessels were made from S-glass, graphite, and Kevlar 49 fibers and were equipped with both structural and nonstructural liners. Pressure vessel structural efficiencies were attained which represented weight savings, using different liners, of 40 to 60 percent over all-titanium pressure vessels. Significant findings in each area are summarized including data from current NASA-Lewis Research Center contractual and in-house programs.

  12. Nonlinear finite element analysis of mechanical characteristics on CFRP composite pressure vessels

    NASA Astrophysics Data System (ADS)

    Liu, Dong-xia; Liang, Li; Li, Ming

    2010-06-01

    CFRP(Carbon Fibre Reinforced Plastic) composite pressure vessel was calculated using finite element program of ANSYS for their mechanical characteristics in this paper. The elastic-plastic model and elements of Solid95 were selected for aluminium alloys of gas cylinder. Also liner-elastic model and layer elements of Shell99 were adopted for carbon fibre/epoxy resin. The stress state of CFRP composite pressure vessel was calculated under different internal pressures include pre-stressing pressures, working pressures, test hydraulic pressures, minimum destructive pressures etcetera to determine the size of gas cylinder and layer parameter of carbon fibre. The mechanical characteristics CFRP composite vessel could were using to design and test of gas cylinder. Numerical results showed that finite element model and calculating method were efficient for study of CFRP gas cylinder and useful for engineering design.

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

    NASA Technical Reports Server (NTRS)

    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.

  14. Effects of shear forces and pressure on blood vessel function and metabolism in a perfusion bioreactor.

    PubMed

    Hoenicka, Markus; Wiedemann, Ludwig; Puehler, Thomas; Hirt, Stephan; Birnbaum, Dietrich E; Schmid, Christof

    2010-12-01

    Bovine saphenous veins (BSV) were incubated in a perfusion bioreactor to study vessel wall metabolism and wall structure under tissue engineering conditions. Group 1 vessels were perfused for 4 or 8 days. The viscosity of the medium was increased to that of blood in group 2. Group 3 vessels were additionally strained with luminal pressure. Groups 1-d through 3-d were similar except that BSV were endothelium-denuded before perfusion. Groups 1-a through 3-a used native vessels at elevated flow rates. Group 3 vessels responded significantly better to noradrenaline on day 4, whereas denuded vessels showed attenuated responses (p < 0.001). Tetrazolium dye reduction did not depend on perfusion conditions or time except for denuded vessels. pO₂ gradients across the vessels were independent of time and significantly higher in group 2 (p < 0.001). BSV converted glucose stoichiometrically to lactate except vessels of groups 3, 1-d, and 3-d which released more lactate than glucose could supply (p < 0.001). Group 1 vessels as well as all vessels perfused with elevated flow rates showed a loss of endothelial cells after 4 days, whereas group 2 and 3 vessels retained most of the endothelium. These data suggest that vessel metabolism was not limited by oxygen supply. Shear forces did not affect glucose metabolism but increased oxygen consumption and endothelial cell survival. Luminal pressure caused the utilization of energy sources other than glucose, as long as the endothelium was intact. Therefore, vessel metabolism needs to be monitored during tissue engineering procedures which challenge the constructs with mechanical stimuli.

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

    SciTech Connect

    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.

  16. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 196.30-1 Section 196.30-1 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) OCEANOGRAPHIC RESEARCH VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-1 Repairs to boilers and...

  17. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    NASA Technical Reports Server (NTRS)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

    To provide NASA with a suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for the vessels described above, Southwest Research Institute® (SwRI®) was contracted in two phases to obtain relevant material property data from a representative vessel. An initial characterization of the strength, fracture and fatigue crack growth properties was performed in Phase 1. Based on the results and recommendations of Phase 1, a more extensive material property characterization effort was developed in this Phase 2 effort. This Phase 2 characterization included additional strength, fracture and fatigue crack growth of the multilayer vessel and head materials. In addition, some more limited characterization of the welds and heat affected zones (HAZs) were performed. This report

  18. High Pressure Composite Overwrapped Pressure Vessel (COPV) Development Tests at Cryogenic Temperatures

    NASA Technical Reports Server (NTRS)

    Ray, David M.; Greene, Nathanael J.; Revilock, Duane; Sneddon, Kirk; Anselmo, Estelle

    2008-01-01

    Development tests were conducted to evaluate the performance of 2 COPV designs at cryogenic temperatures. This allows for risk reductions for critical components for a Gaseous Helium (GHe) Pressurization Subsystem for an Advanced Propulsion System (APS) which is being proposed for NASA s Constellation project and future exploration missions. It is considered an advanced system since it uses Liquid Methane (LCH4) as the fuel and Liquid Oxygen (LO2) as the oxidizer for the propellant combination mixture. To avoid heating of the propellants to prevent boil-off, the GHe will be stored at subcooled temperatures equivalent to the LO2 temperature. Another advantage of storing GHe at cryogenic temperatures is that more mass of the pressurized GHe can be charged in to a vessel with a smaller volume, hence a smaller COPV, and this creates a significant weight savings versus gases at ambient temperatures. The major challenge of this test plan is to verify that a COPV can safely be used for spacecraft applications to store GHe at a Maximum Operating Pressure (MOP) of 4,500 psig at 140R to 160R (-320 F to -300 F). The COPVs for these tests were provided by ARDE , Inc. who developed a resin system to use at cryogenic conditions and has the capabilities to perform high pressure testing with LN2.

  19. Filament-reinforced metal composite pressure vessel evaluation and performance demonstration

    NASA Technical Reports Server (NTRS)

    Landes, R. E.

    1976-01-01

    Two different Kevlar-49 filament-reinforced metal sphere designs were developed, and six vessels of each type were fabricated and subjected to fatigue cycling, sustained loading, and hydrostatic burst. The 61 cm (24 inch) diameter Kevlar-49/cryoformed 301 stainless steel pressure vessels demonstrated the required pressure cycle capability, burst factor of safety, and a maximum pressure times volume divided by weight (pV/W) performance of 210 J/g (834 000 in-lb/lbm) at burst; this represented a 25 to 30% weight saving over the lightest weight comparable, 6A1-4V Ti, homogeneous pressure vessel. Both the Kevlar/stainless steel design and the 97 cm (38 inch) diameter Kevlar-49/2219-T62 aluminum sphere design demonstrated nonfragmentation and controlled failure mode features when pressure cycled to failure at operating pressure. When failure occurred during pressure cycling, the mode was localized leakage and not catastrophic. Kevlar/stainless steel vessels utilized a unique conical boss design, and Kevlar/aluminum vessels incorporated a tie-rod to carry port loads; both styles of polar fittings performed as designed during operational testing of the vessels.

  20. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOEpatents

    Ekeroth, D.E.; Orr, R.

    1993-12-07

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel. 5 figures.

  1. Nuclear reactor having a polyhedral primary shield and removable vessel insulation

    DOEpatents

    Ekeroth, Douglas E.; Orr, Richard

    1993-01-01

    A nuclear reactor is provided having a generally cylindrical reactor vessel disposed within an opening in a primary shield. The opening in the primary shield is defined by a plurality of generally planar side walls forming a generally polyhedral-shaped opening. The reactor vessel is supported within the opening in the primary shield by reactor vessel supports which are in communication and aligned with central portions of some of the side walls. The reactor vessel is connected to the central portions of the reactor vessel supports. A thermal insulation polyhedron formed from a plurality of slidably insertable and removable generally planar insulation panels substantially surrounds at least a portion of the reactor vessel and is disposed between the reactor vessel and the side walls of the primary shield. The shape of the insulation polyhedron generally corresponds to the shape of the opening in the primary shield. Reactor monitoring instrumentation may be mounted in the corners of the opening in the primary shield between the side walls and the reactor vessel such that insulation is not disposed between the instrumentation and the reactor vessel.

  2. Adaptation of mesenteric lymphatic vessels to prolonged changes in transmural pressure

    PubMed Central

    Dongaonkar, R. M.; Nguyen, T. L.; Hardy, J.; Laine, G. A.; Wilson, E.; Stewart, R. H.

    2013-01-01

    In vitro studies have revealed that acute increases in transmural pressure increase lymphatic vessel contractile function. However, adaptive responses to prolonged changes in transmural pressure in vivo have not been reported. Therefore, we developed a novel bovine mesenteric lymphatic partial constriction model to test the hypothesis that lymphatic vessels exposed to higher transmural pressures adapt functionally to become stronger pumps than vessels exposed to lower transmural pressures. Postnodal mesenteric lymphatic vessels were partially constricted for 3 days. On postoperative day 3, constricted vessels were isolated, and divided into upstream (UP) and downstream (DN) segment groups, and instrumented in an isolated bath. Although there were no differences between the passive diameters of the two groups, both diastolic diameter and systolic diameter were significantly larger in the UP group than in the DN group. The pump index of the UP group was also higher than that in the DN group. In conclusion, this is the first work to report how lymphatic vessels adapt to prolonged changes in transmural pressure in vivo. Our results suggest that vessel segments upstream of the constriction adapt to become both better fluid conduits and lymphatic pumps than downstream segments. PMID:23666672

  3. Adaptation of mesenteric lymphatic vessels to prolonged changes in transmural pressure.

    PubMed

    Dongaonkar, R M; Nguyen, T L; Quick, C M; Hardy, J; Laine, G A; Wilson, E; Stewart, R H

    2013-07-15

    In vitro studies have revealed that acute increases in transmural pressure increase lymphatic vessel contractile function. However, adaptive responses to prolonged changes in transmural pressure in vivo have not been reported. Therefore, we developed a novel bovine mesenteric lymphatic partial constriction model to test the hypothesis that lymphatic vessels exposed to higher transmural pressures adapt functionally to become stronger pumps than vessels exposed to lower transmural pressures. Postnodal mesenteric lymphatic vessels were partially constricted for 3 days. On postoperative day 3, constricted vessels were isolated, and divided into upstream (UP) and downstream (DN) segment groups, and instrumented in an isolated bath. Although there were no differences between the passive diameters of the two groups, both diastolic diameter and systolic diameter were significantly larger in the UP group than in the DN group. The pump index of the UP group was also higher than that in the DN group. In conclusion, this is the first work to report how lymphatic vessels adapt to prolonged changes in transmural pressure in vivo. Our results suggest that vessel segments upstream of the constriction adapt to become both better fluid conduits and lymphatic pumps than downstream segments.

  4. Pressure Vessel and Internals of the International Reactor Innovative and Secure

    SciTech Connect

    Lombardi, C.V.; Padovani, E.; Cammi, A.; Collado, J.M.; Santoro, R.T.; Barnes, J.M.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral light water cooled, low-to-medium power reactor, which addresses the requirements defined by the US DOE for Generation IV reactors. Its integrated layout features a pressure vessel containing all the main primary circuit components: the internals and the biological shield, here described together with the pressure vessel, plus the steam generators, the pressurizer, and the main coolant pumps described in companion papers. For this reason the pressure vessel is a crucial component of the plant, which deserves the most demanding design effort. The wide inner annulus around the core is exploited to insert steel plates, in order to improve the inner shielding capability up to the elimination of the external biological shielding and to simplify decommissioning activities by having all the irradiated components inside the vessel. (authors)

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

    NASA Technical Reports Server (NTRS)

    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.

  6. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ...) NAUTICAL SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-5 Inspection of boilers, pressure... (Marine Engineering) of this chapter, insofar as they relate to tests and inspection of cargo vessels. ...

  7. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ...) NAUTICAL SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-5 Inspection of boilers, pressure... (Marine Engineering) of this chapter, insofar as they relate to tests and inspection of cargo vessels. ...

  8. 98. ARAIII. ML1 reactor pressure vessel is lowered into reactor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    98. ARA-III. ML-1 reactor pressure vessel is lowered into reactor pit by hoist. July 13, 1963. Ineel photo no. 63-4049. Photographer: Lowin. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

    NASA Technical Reports Server (NTRS)

    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. Influence and Modeling of Residual Stresses in Thick Walled Pressure Vessels with Through Holes

    DTIC Science & Technology

    2012-02-28

    Technical Report ARWSB-TR-12003 INFLUENCE AND MODELING OF RESIDUAL STRESSES IN THICK WALLED PRESSURE VESSELS WITH...DATE (DD-MM-YYYY) 28/02/2012 2. REPORT TYPE Technical Report 3. DATES COVERED (From - To) 4. TITLE AND SUBTITLE Influence and Modeling of...Residual Stresses in Thick Walled Pressure Vessels with Through Holes 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c. PROGRAM ELEMENT

  11. ‘Sausage-string’ deformations of blood vessels at high blood pressures

    NASA Astrophysics Data System (ADS)

    Alstrøm, P.; Mikkelsen, R.; Gustafsson, F.; Holstein-Rathlou, N.-H.

    1999-12-01

    A new instability is proposed to explain the ‘sausage-string’ patterns of alternating constrictions and dilatations formed in blood vessels at high blood pressure conditions. Our theory provides predictions for the conditions under which the cylindrical geometry of a blood vessel becomes unstable. The theory is related to experimental observations in rats, where high blood pressure is induced by intravenous infusion of angiotensin II.

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

    NASA Astrophysics Data System (ADS)

    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

  13. Safety-Valve Mechanism For Pressure-Vessel Window

    NASA Technical Reports Server (NTRS)

    Mccoomb, E. J.

    1994-01-01

    Pressure-activated valve mechanism seals small window in pressure chamber if window cracks or breaks, thereby preventing continued leakage or sudden decompression. Window used in experiments involving optical observation of processes in chamber. Valve mechanism activated by pressure from gas leaking through window.

  14. Fracture analysis of surface and through cracks in cylindrical pressure vessels

    NASA Technical Reports Server (NTRS)

    Newman, J. C., Jr.

    1976-01-01

    A previously developed fracture criterion was applied to fracture data for surface- and through-cracked cylindrical pressure vessels to see how well the criterion can correlate fracture data. Fracture data from the literature on surface cracks in aluminum alloy, steel, and epoxy vessels, and on through cracks in aluminum alloy, titanium alloy steel, and brass vessels were analyzed by using the fracture criterion. The criterion correlated the failure stresses to within + or - 10 percent for either surface or through cracks over a wide range of crack size and vessel diameter. The fracture criterion was also found to correlate failure stresses to within + or - 10 percent for flat plates (center-crack or double-edge-crack tension specimens) and cylindrical pressure vessels containing through cracks.

  15. Fracture analysis of surface and through-cracks in cylindrical pressure vessels

    NASA Technical Reports Server (NTRS)

    Newman, J. C., Jr.

    1976-01-01

    A previously developed fracture criterion was applied to surface- and through-cracked cylindrical pressure vessels to see how well the criterion can correlate fracture data. Fracture data from the literature on surface cracks in aluminum alloy, steel, and epoxy vessels and on through cracks in aluminum alloy, titanium alloy, steel, and brass vessels were analyzed using the fracture criterion. The criterion correlated the failure stresses to within + or - 10 percent for either surface or through cracks over a wide range of crack size and vessel diameter. The fracture criterion was also found to correlate failure stresses from flat plates (center-crack or double-edge-crack tension specimens) and cylindrical pressure vessels containing through - cracks within + or - 10 percent.

  16. Mechanical Behavior of A Metal Composite Vessels Under Pressure At Cryogenic Temperatures

    NASA Astrophysics Data System (ADS)

    Tsaplin, A. I.; Bochkarev, S. V.

    2016-01-01

    Results of an experimental investigation into the deformation and destruction of a metal composite vessel with a cryogenic gas are presented. Its structure is based on basalt, carbon, and organic fibers. The vessel proved to be serviceable at cryogenic temperatures up to a burst pressure of 45 MPa, and its destruction was without fragmentation. A mathematical model adequately describing the rise of pressure in the cryogenic vessel due to the formation of a gaseous phase upon boiling of the liquefied natural gas during its storage without drainage at the initial stage is proposed.

  17. Pressure Vessel with Impact and Fire Resistant Coating and Method of Making Same

    NASA Technical Reports Server (NTRS)

    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 hear absorption.

  18. Pressure Vessel with Impact and Fire Resistant Coating and Method of Making Same

    NASA Technical Reports Server (NTRS)

    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 hear absorption.

  19. Structural considerations in design of lightweight glass-fiber composite pressure vessels

    NASA Technical Reports Server (NTRS)

    Faddoul, J. R.

    1973-01-01

    The design concepts used for metal-lined glass-fiber composite pressure vessels are described, comparing the structural characteristics of the composite designs with each other and with homogeneous metal pressure vessels. Specific design techniques and available design data are identified. The discussion centers around two distinctly different design concepts, which provide the basis for defining metal lined composite vessels as either (1) thin-metal lined, or (2) glass fiber reinforced (GFR). Both concepts are described and associated development problems are identified and discussed. Relevant fabrication and testing experience from a series of NASA-Lewis Research Center development efforts is presented.

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

    NASA Technical Reports Server (NTRS)

    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.