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Sample records for concrete pressure-vessel subjected

  1. Behaviour of a pre-stressed concrete pressure-vessel subjected to a high temperature gradient

    International Nuclear Information System (INIS)

    Dubois, F.

    1965-01-01

    After a review of the problems presented by pressure-vessels for atomic reactors (shape of the vessel, pressures, openings, foundations, etc.) the advantages of pre-stressed concrete vessels with respect to steel ones are given. The use of pre-stressed concrete vessels however presents many difficulties connected with the properties of concrete. Thus, because of the absence of an exact knowledge of the material, it is necessary to place a sealed layer of steel against the concrete, to have a thermal insulator or a cooling circuit for limiting the deformations and stresses, etc. It follows that the study of the behaviour of pre-stressed concrete and of the vessel subjected- to a high temperature gradient can yield useful information. A one-tenth scale model of a pre-stressed concrete cylindrical vessel without any side openings and without a base has been built. Before giving a description of the tests the authors consider some theoretical aspects concerning 'scale model-actual structure' similitude conditions and the calculation of the thermal and mechanical effects. The pre-stressed concrete model was heated internally by a 'pyrotenax' element and cooled externally by a very strong air current. The concrete was pre-stressed using horizontal and vertical cables held at 80 kg/cm 2 ; the thermal gradient was 160 deg. C. During the various tests, measurements were made of the overall and local deformations, the changes in water content, the elasticity modulus, the stress and creep of the cables and the depths of the cracks. The overall deformations observed are in line with thermal deformation theories and the creep of the cables attained 20 to 30 per cent according to their position relative to the internal surface. The dynamic elasticity modulus decreased by half but the concrete keeps its good mechanical properties. Finally, cracks 8 to 12 cm deep and 2 to 3 mms wide appeared in that part of the concrete which was not pre-stressed. The results obtained make it

  2. Model tests for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Stoever, R.

    1975-01-01

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

  3. Method of detecting construction faults in concrete pressure vessels

    International Nuclear Information System (INIS)

    Robertson, S.A.; Duhoux, M.; Dawance, G.; Carrie, C.; Morel, D.

    1976-01-01

    A major problem in the design and construction of concrete pressure vessels for nuclear power stations is the risk of excessive air leaks through the concrete itself, due to faulty construction. The 'sonic coring' method of non-destructive concrete testing has been used successfully in pile and diaphragm wall construction control for several years, and the potential use of this method to control the presence of faults in concrete pressure vessels is here described. (author)

  4. Reinforced-concrete pressure vessel for a nuclear power plant

    International Nuclear Information System (INIS)

    Schwiers, H.G.; Schoening, J.

    1979-01-01

    The gas turbo-engine of the THTR-300 is installed in a horizontal duct of the prestressed concrete pressure vessel. The cavern and its recesses for the steam generators are arranged above and laterally away from this duct. By this means prestressing of the individual regions of the pressure vessel may be adapted to the pressure existing in the different cavities. (RW) [de

  5. Design criteria for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1989-01-01

    The work concerned with the PCRVs has been focussed on topics which are not sufficiently covered by the usual codes with respect to the special structure of PCRVs and the special demands on it, and different investigations yielding a basis for such specific design criteria have been carried out. Only a couple of subjects being in the fore under the aspect of defining quality enlarging design criteria for PCRVs are outlined. The materials for the concrete to be used for the PCRVs are carefully selected. (DG)

  6. Cylindrical prestressed concrete pressure vessel for a nuclear power plant

    International Nuclear Information System (INIS)

    Horner, M.; Hodzic, A.; Haferkamp, D.

    1976-01-01

    A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

  7. Prestressed concrete pressure vessel for a nuclear reactor

    International Nuclear Information System (INIS)

    Ritz, L.

    1976-01-01

    A prestressed concrete pressure vessel is described which is to be used for a nuclear reactor. It is of integrated construction, has a fixed amount of prestressing and can be used at very high internal pressures. The re-inforcement is arranged in U shapes in the concrete and the space of the reactor core and equipment is situated between the two uprights (of the U). Their ends are anchored to the outer walls of the concrete. All ends of the cross braced re-inforcement are equally distributed over the circumference of the common concrete structure along its height. (UWI) [de

  8. In-service supervision of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Zemann, H.; Mayer, N.; Amberg, C.

    1985-01-01

    On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C/50 bar). (Author)

  9. In-service supervision of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Zemann, H.; Weissbacher, L.; Mayer, N.; Amberge, C.

    1985-01-01

    On-line measurements of the physical state of a prestressed concrete pressure vessel, and comparison with the design predictions of the distribution of temperature, strain and stress within the concrete member and the criteria of layout, provide an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed with reference to the prototype vessel at Seibersdorf Research Centre during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C, 50 bar). (author)

  10. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Berg, S.; Loeseth, S.; Holand, I.

    1977-01-01

    A computational model for circular symmetric reinforced concrete shell problems is described. The model is based on the Finite Element Method. Non-linear stress-strain constitutive relations are used for the concrete, the reinforcement and for the liner. The reinforcement layers may be of different steel qualities. Each layer may be given a specified prestressing. This can be done at the beginning of the computations or the specific reinforcement layer can be considered inactive until a specified level of loading is reached. Thus, the prestressing procedure may also be analyzed in detail. Bond-slip effects are not accounted for. However, no bond may be assumed for prestressing cables by inserting special reinforcement elements. Several models of prestressed concrete reactor pressure vessels which have been tested up to rupture have been analysed. Analytical (numerical) models for reinforced concrete are also discussed on a more general basis. (Auth.)

  11. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

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

    International Nuclear Information System (INIS)

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

    The purpose of this investigation was to determine the corrosion behavior of a high strength steel (ASTM A416-74 grade 270), typical of those used as tensioning tendons in prestressed concrete pressure vessels, in several corrosive environments and to demonstrate the protection afforded by coating the steel with either of two commercial petroleum-base greases or Portland Cement grout. In addition, the few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors are 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 but small flaws in the grease coatings were detrimental; flaws or cracks less than 1 mm wide in the grout were without effect

  13. Cylindrical reinforced-concrete pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Vaessen, F.

    1975-01-01

    The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

  14. Minimum weight design of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Boes, R.

    1975-01-01

    A method of non-linear programming for the minimization of the volume of rotationally symmetric prestressed concrete reactor pressure vessels is presented. It is assumed that the inner shape, the loads and the degree of prestressing are prescribed, whereas the outer shape is to be detemined. Prestressing includes rotational and vertical tension. The objective function minimizes the weight of the PCRV. The constrained minimization problem is converted into an unconstrained problem by the addition of interior penalty functions to the objective function. The minimum is determined by the variable metric method (Davidson-Fletcher-Powell), using both values and derivatives of the modified objective function. The one-dimensional search is approximated by a method of Kund. Optimization variables are scaled. The method is applied to a pressure vessel like for THTR. It is found that the thickness of the cylindrical wall may be reduced considerably for the load cases considered in the optimization. The thickness of the cover is reduced slightly. The largest reduction in wall thickness occurs at the junction of wall and cover. (Auth.)

  15. State-of-the-art and prospets for designing and constraction of prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Short review of reports submitted to the symposium on pressure vessels, which was conducted in Calgary (Canada), has been presented. New tendencies of designing of prestressed concrete pressure vessels (PCPV) for nuclear for nuclear reactors are noted. Construction of hot vessel liner is studied. A conclusion is drawn on prospects of PCPV creation

  16. An introduction to the analysis of multi-cavity prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Silva, M.C.A.T. da.

    1986-01-01

    The present work is a study of multi-cavity prestressed concrete pressure vessels (PCRV) for nuclear reactors. A review is made of the designs, analises and models of multi-cavity concrete pressure vessels. A preliminary evaluation of the NONSAP program for applications in complex three-dimensional structures such as a multi-cavity pressure vessel is also made. A model of a PCRV of a 1000 MW(e) high-temperature gas cooled reactor was selected for a three-dimensional analysis with the NONSAP program. The results obtained are compared with experimental data. (Author) [pt

  17. UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

    Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

  18. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

    Numerical procedures for predicting the nonlinear behaviour of a prestressed concrete reactor vessel over its design life are discussed. The numerical models are constructed by combining three-dimensional isoparametric finite elements which simulate the concrete, thin shell elements which simulate steel liner plates and layers of reinforcement steel, and axial elements for discrete prestressing cables. Nonlinearity under compressive stress, multi-dimensional cracking, shrinkage, and stress/temperature induced creep of concrete are considered in addition to the elastic plastic behaviour of the liner and reinforcing steel. The analysis of an actual PCRV is described. Stress contours and cracking patterns in the region of cutouts corresponding to operational pressure and temperature loads are illustrated. The effects of creep, unloading,and creep recovery are then shown. Lastly, a strategy for assessing the performance over its design life is discussed. (Auth.)

  19. Reactors with pressure vessel in pre-stressed concrete

    International Nuclear Information System (INIS)

    Devillers, Christian; Lafore, Pierre

    1964-12-01

    After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

  20. Feasibility of Rectangular Concrete Pressure Vessels for Human Occupancy

    Science.gov (United States)

    1990-07-01

    work this stiff mix a superplasticizer was added. Predictions are now being made of 25,000 to 40,000 psi concrete in the 1990s (ref. 10). High...Vibration should be required and vibration times recorded. If vibration is not practicable, superplasticizers PAGE -22- (water reducing agents) should be used

  1. The dynamic relaxation method in the structural analysis of concrete pressure vessels

    International Nuclear Information System (INIS)

    Davidson, I.; Assis Bastos, M.R. de; Camargo, P.B. de.

    1977-01-01

    The dynamic relaxation method, applied to 3 dimensional concrete structures, especially pressure vessels, is demonstrated. It utilizes the finite difference method and allows the growth of cracks to be followed up to the point of vessel rupture. A FORTRAN IV program is developed, which can also be utilized, with the necessary modifications, for other structure calculations [pt

  2. Experience of in-service surveillance and monitoring of prestressed concrete pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    Irving, J.; Smith, J.R.; Eadie, D.McD.; Hornby, I.W.

    1976-01-01

    Details are given of the statutory requirements for the inspection of prestressed concrete pressure vessels in the United Kingdom, with particular emphasis on the prestressing system. The results of periodic examinations under the Licencing Conditions of the Oldbury and Wylfa vessels are presented and discussed in relation to design expectations and future requirements. Strain, moisture and temperature records obtained from the Oldbury PCPV's over a 10 year period are compared with prediction and new developments in vessel instrumentation are discussed. (author)

  3. Monitoring of prestressed concrete pressure vessels. II. performance of selected concrete embedment strain meters under normal and extreme environmental conditions

    International Nuclear Information System (INIS)

    Naus, D.J.; Hurtt, C.C.

    1978-10-01

    Unique types of instrumentation are used in prestressed concrete pressure vessels (PCPVs) to measure strains, stresses, deflections, prestressing forces, moisture content, temperatures, and possibly cracking. Their primary purpose is to monitor these complex structures throughout their 20- to 30-year operating lifetime in order to provide continuing assurance of their reliability and safety. Numerous concrete embedment instrumentation systems are available commercially. Since this instrumentation is important in providing continuing assurance of satisfactory performance of PCPVs, the information provided must be reliable. Therefore, laboratory studies were conducted to evaluate the reliability of these commercially available instrumentation systems. This report, the second in a series related to instrumentation embedded in concrete, presents performance-reliability data for 13 types of selected concrete embedment strain meters which were subjected to a variety of loading environments, including unloaded, thermally loaded, simulated PCPV, and extreme environments. Although only a limited number of meters of each type were tested in any one test series, the composite results of the investigation indicate that the majority of these meters would not be able to provide reliable data throughout the 20- to 30-year anticipated operating life of a PCPV. Specific conclusions drawn from the study are: (1) Improved corrosion-resistant materials and sealing techniques should be developed for meters that are to be used in PCPV environments. (2) There is a need for the development of meters that are capable of surviving in concretes where temperatures in excess of 66 0 C are present for extended periods of time. (3) Research should be conducted on other measurement techniques, such as inductance, capacitance, and fluidics

  4. Calculation of Prestressed Pressure Vessel Taking into Account the Concrete Temperature Inhomogeneity

    Science.gov (United States)

    Andreev, Vladimir

    2018-03-01

    The paper deals with the problem of determining the stress state of the pressure vessel (PV) with considering the concrete temperature inhomogeneity. Such structures are widely used in heat power engineering, for example, in nuclear power engineering. The structures of such buildings are quite complex and a comprehensive analysis of the stress state in them can be carried out either by numerical or experimental methods. However, a number of fundamental questions can be solved on the basis of simplified models, in particular, studies of the effect on the stressed state of the inhomogeneity caused by the temperature field.

  5. Review of current practices and requirements for the inspection of prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Reimann, K.J.

    1980-12-01

    Code requirements for pre- and in-service inspection of prestressed concrete pressure vessels as utilized in gas-cooled reactors are reviewed and compared with practices and experiences during construction, commissioning, and operation of such reactors. The pre-service inspection relies heavily on embedded instrumentation for measurements of stresses, temperatures, and displacements. The same instrumentation is later used for in-service surveillance, which additionally includes visual examination of exposed surfaces, monitoring of tendon conditions, and measurement of tendon loads. Improvement of present monitoring instrumentation and/or techniques, rather than development of new in-service inspection methods, is recommended

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

    International Nuclear Information System (INIS)

    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

  7. Experimental analysis of a nuclear reactor prestressed concrete pressure vessels model

    International Nuclear Information System (INIS)

    Vallin, C.

    1980-01-01

    A comprehensible analysis was made of the performance of each set of sensors used to measure the strain and displacement of a 1/20 scale Prestressed Concrete Pressure Vessel (PCPV) model tested at the Instituto de Pesquisas Energeticas e Nucleares (IPEN). Among the three Kinds of sensors used (strain gage, displacement transducers and load cells) the displacement transducers showed the best behavior. The displacemente transducers data was statistically analysed and a linear behavior of the model was observed during the first pressurizations tests. By means of a linear statistical correlation between experimental and expected theoretical data it was found that the model looses the linearity at a pressure between 110-125 atm. (Author) [pt

  8. Process for producing curved surface of membrane rings for large containers, particulary for prestressed concrete pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1977-01-01

    Membrane rings for large pressure vessels, particularly for prestressed-concrete pressure vessels, often have curved surfaces. The invention describes a process of producing these at site, which is particularly advantageous as the forming and installation of the vessel component coincide. According to the invention, the originally flat membrane ring is set in a predetermined position, is then pressed in sections by a forming tool (with a preformed support ring as opposite tool), and shaped. After this, the shaped parts are welded to the ring-shaped wall parts of the large vessel. The manufacture of single and double membrane rings arrangements is described. (HP) [de

  9. An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

    International Nuclear Information System (INIS)

    1982-04-01

    An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Naus, D.J.

    1981-05-01

    Prestressed concrete pressure vessels (PCPVs) are designed to serve as primary pressure containment structures. The safety of these structures depends on a correct assessment of the loadings and proper design of the vessels to accept these loadings. Proper vessel design requires a knowledge of the component (material) properties. Because concrete is one of the primary constituents of PCPVs, knowledge of its behavior is required to produce optimum PCPV designs. Concrete material systems are reviewed with respect to constituents, mix design, placing, curing, and strength evaluations, and typical concrete property data are presented. Effects of extreme loadings (elevated temperature, multiaxial, irradiation) on concrete behavior are described. Finally, specialty concrete material systems (high strength, fibrous, polymer, lightweight, refractory) are reviewed. 235 references.

  11. BBRV post-tensioning systems as applied to reactor containments and prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Thorpe, W.; Speck, F.E.

    1976-01-01

    Nuclear containments and pressure vessels can be post-tensioned by using two basically different methods: tendons and winding. The fundamental differences between the two concepts are shown by introductory examples. A discussion of tendon units, usually lying in the range 4000 to 10,000 kN, is followed by a detailed presentation of the BBRV winding system. After giving a short comment to factors influencing the choice of a post-tensioning system the authors discuss specific aspects of some application groups: cable layout with containments and pressure vessels, conditions for a wrapped design, corrosion protection. (author)

  12. Meeting 'Prestressed-concrete reactor pressure vessels', 13th and 14th october 1975, Berlin

    International Nuclear Information System (INIS)

    Schickert, G.

    1976-01-01

    Influence of radioactive radiation on the mechanical properties of concrete; behaviour of concrete in short-time testing under multiaxial mechanical stresses; behaviour of concrete in long-time testing under multiaxial mechanical stresses at higher temperatures; temperature stress of concrete; strength formation of concrete; steel fiber concrete. (LH) [de

  13. Suspension of a steam raising unit in a reinforced concrete pressure vessel of a high temperature reactor

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.; Stracke, W.

    1982-01-01

    The invention concerns a supporting device in a steam raising unit, which is situated in the reinforced concrete pressure vessel of a high temperature reactor. The supporting device is situated in the space between the steam raising unit skirt and the steam raising unit jacket. If the suspension of the steam raising unit skirt fails, the steam raising unit skirt is well fixed by the supporting device in the steam raising unit jacket, without the feed water and steam pipes in the upper part of the steam raising unit being damaged. (orig.) [de

  14. The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Pin-Chiun [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China); Chou, Hsoung-Wei, E-mail: hwchou@iner.gov.tw [Institute of Nuclear Energy Research, Taoyuan 32546, Taiwan, ROC (China); Ferng, Yuh-Ming [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China)

    2016-02-15

    Highlights: • Probabilistic fracture mechanics method was used to analyze a reactor pressure vessel. • Effects of copper and nickel contents on RPV fracture probability under PTS were investigated and discussed. • Representative PTS transients of Beaver Valley nuclear power plant were utilized. • The range of copper and nickel contents of the RPV materials were suggested. • With different embrittlement levels the dominated PTS category is different. - Abstract: The radiation embrittlement behavior of reactor pressure vessel shell is influenced by the chemistry concentration of metal materials. This paper aims to study the effects of copper and nickel content variations on the fracture risk of pressurized water reactor (PWR) pressure vessel subjected to pressurized thermal shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel assumed with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the U.S. NRC's new PTS rule are applied as the loading condition. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

  15. Principles of design and construction for the top caps of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Hughes, A.N.; Bellwood, G.N.; Paton, A.A.

    1976-01-01

    The building of the top cap poses problems because of the number of penetrations to be cast therein. The fuel and control system routes need to be tightly specified and controlled so that during station life misalignments do not occur which interfere with the fuelling and control operations. The paper outlines the route requirements and illustrates how these affect the tolerances and movements which can be allowed at various stages of construction. Development work is discussed to show the necessity of resolving the different priorities of design, programme and overall pressure vessel construction requirements, so that the reactor build is not inhibited by the special demands of the top cap, and the integration of the monitoring and survey systems during the top cap build are explained. (author)

  16. Design and construction of the prestressed concrete boiler closures for the Hartlepool and Heysham pressure vessels

    International Nuclear Information System (INIS)

    Crowder, R.; Howells, R.M.; Paton, A.A.

    1976-01-01

    At a relatively late stage in the station design, the boiler closures for the reactor vessels at Hartlepool and Heysham were changed from steel to prestressed concrete. This paper sets out the criteria which were finally evolved for the new style of closure and describes the way in which the prestressed concrete closure's parts were designed to satisfy these criteria. With both the civil and mechanical components of the closure having their own specific requirements, close co-operation was necessary between these disciplines to ensure that a compatible and practical closure design resulted. This close interrelationship has been carried through into the construction stage and a special concreting and prestressing factory has been built adjacent to the works of the mechanical component fabricator. This enabled an optimum manufacturing cycle to be followed and the important aspects of this are described in the paper. (author)

  17. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

  18. Design criteria for prestressed concrete pressure vessels for high temperature reactors

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1991-01-01

    This paper summarizes the work on design criteria for concrete structures of Prestressed Concrete Reactor Vessels (PCRVs), which has been carried out since 1984 by a couple of competent institutions. After some basic considerations on the safety demands on PCRVs, especially their Prestressed Concrete Structure (PCS), and the consequences for an elevated level of quality to be ensured by the design criteria, an impression is given, first, by what means a higher quality standard is gained with respect to selection of materials and specification of material data in comparison to the usual building industry and what kind of criteria on this behalf should be fixed in a PCRV code. As a further quality increasing feature, the specific demands on design analysis as practised according to the present state of science and as to be treated within a code are discussed. This concerns analyses for steady state and transient temperatures as well as stress and strain analyses for service and ultimate load conditions. It is outlined to what degree calculation models should be detailed, which includes statements about admissible idealizations. As a central topic the question is discussed in what way the ultimate load capacity has to be evaluated, thereby presenting results of some investigations pointing out the conditions under which the design is determined by the different kinds of ultimate load conditions. Finally, some reflections on the demands on monitoring the PCS behaviour during its lifetime and on several questions still to be answered in this field are expressed. (orig.)

  19. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Suzuki, K.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  20. FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

    Full Text Available ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN

  1. Method of producing the arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1976-01-01

    In producing arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels for nuclear power plants, it is of advantage to manufacture these directly on the construction site. According to the invention the, at first level, diaphragm ring is put on the predetermined place, sectionally pressed against and shaped by a shaping tool - with a profiled supporting ring as a counter-acting tool - and afterwards welded together with the annular wall sections of the large container along the shaped parts. The manufacture of single and double configurations of diaphragm rings is described. It is of advantage if shaping and mounting position coincide. (UWI) [de

  2. Design and construction of a prestressed concrete pressure vessel for a working pressure of 69N/mm2 (10,000 p.s.i)

    International Nuclear Information System (INIS)

    Dawson, P.

    1977-01-01

    Construction is nearing completion of a pressure vessel with a chamber 9.15 m (30 ft.) high and 3.05 m (10 ft.) internal diameter for hydraulic tests on marine components up to 69 N/mm 2 (10,000 p.s.i.) working pressure. The chamber comprises a steel cylinder, with independent end plates contained within a prestressed concrete structure. The cylinder is constructed in two halves, each consisting of three forged rings, 170 mm thick, shrink-fitted onto a 90 mm thick liner. It rests on a 100 mm thick bottom plate, provided with a band of hard-facing overlay on which the cylinder slides in response to changes of test medium pressure. Models to be tested within the chamber are hung from a removeable 150 mm thick top plate. A central elliptical hatch provides access into the chamber. Special sealing assemblies are fitted at the junction of the cylinder sections and between the cylinder and end plates. These seals are capable of accepting radial expansion of the cylinder and corresponding vertical movements at the upper seal arising from elastic movements of the enclosing structure. The top plate is restrained by a wire-wound prestressed concrete closure plug, itself located by twelve bifurcated inclined steel struts which transfer the load on the top plate into the concrete structure. The struts are retractable to allow removal of the closure plug and top plate. The enclosing concrete structure is 25 m (82 ft.) high and 11 m (36 ft.) diameter. It is vertically prestressed by 180 no. 540 Tonne tendons and circumferentially prestressed by 5 mm wire laid under tension in pre-cast concrete channels by the Taylor Woodrow Wire-Winding System. The structure was analysed, using limit state principles, by computerised elastic and non-elastic dynamic relaxation techniques. The results were evaluated against triaxial stress criteria established from relevant research work and experience obtained from nuclear prestressed concrete pressure vessels

  3. Heating large pressure vessel with a cylindrical cross-section

    International Nuclear Information System (INIS)

    Gabrea, R.; Kaiser, D.; Schwiers, H.G.; Schoening, J.

    1980-01-01

    The proposal concerns the improvement of a construction of the vertical stress system of a prestressed-concrete pressure vessel to take up nuclear power plant components. According to the invention, the vertical bracing, cables are attached outside the pressure-resistant thermal insulation of the pressure container, so that they can be easily checked. They are not subjected to thermal stresses. (UWI) [de

  4. Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks

    International Nuclear Information System (INIS)

    Qian, Guian; Niffenegger, Markus

    2013-01-01

    The reactor pressure vessel (RPV), as one of the most important safety barriers of light water reactors, is exposed to neutron irradiation at elevated temperatures, which results in embrittlement of the RPV steel. One potential challenge to the structural integrity of the RPV in a pressurized water reactor is posed by pressurized thermal shock (PTS). Therefore, the safety of the RPV with regard to neutron embrittlement has to be analyzed. In this paper, the procedure and method for the structural integrity analysis of RPV subjected to PTS are presented. FAVOR and PASCAL, two computer codes widely used for the probabilistic analysis of RPV subjected to PTS, are briefly reviewed and compared. By using FAVOR, a benchmark example is presented to show the procedure and method for the integrity analysis. The influence of warm prestressing (WPS), fracture toughness and constraint effect on the integrity analysis of RPV is discussed. The Master Curve method is more realistic than the ASME model to consider the analysis of fracture toughness and thus is recommended. In order to transfer the fracture toughness data from test specimen to the RPV, local approach provides a probabilistic method

  5. Application of dynamic relaxation and finite elements methods for the structural analysis of a scale model of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Tamura, Masaru

    1979-01-01

    A stress and strain analysis was made of a scale model of a Prestressed Concrete Pressure Vessel for a Boiling Water Reactor. The aim of this work was to obtain an experimental verification of the calculation method actually used at IPEN. The 1/10 scale model was built and tested at the Instituto Sperimentale Modelli e Structture, ISMES, Italy. The dynamic relaxation program PV2-A and the finite element programs , FEAST-1 have been used. A comparative analysis of the final results was made. A preliminary analysis was made for a simplified monocavity model now under development at IPEN with the object of confirming the data and the calculation method used. (author)

  6. 46 CFR 50.30-15 - Class II pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  7. 46 CFR 50.30-20 - Class III pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  8. Nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    McDonald, B.N.

    1976-01-01

    In nuclear power reactor systems which have a reactor core inside a pressure vessel, the feedwater inlet pipe and steam discharge nozzle usually require separate pressure vessel penetrations. This requirement involves a great deal of expensive high quality special machining, welding and weld joint testing. The invention overcomes most of these problems by nestling the feedwater inlet pipe inside the steam discharge nozzle. At the same time the individual heat exchanger modules are supported from the pressure vessel at the same location as the nested feedwater inlet pipe and steam discharge nozzle combination, thus eliminating the need to accomodate troublesome differential thermal expansion problems through special structures within the pressure vessel

  9. Characterizing the effects of elevated temperature on the air void pore structure of advanced gas-cooled reactor pressure vessel concrete using x-ray computed tomography

    Directory of Open Access Journals (Sweden)

    Withers P.J.

    2013-07-01

    Full Text Available X-ray computed tomography (X-ray CT has been applied to nondestructively characterise changes in the microstructure of a concrete used in the pressure vessel structure of Advanced Gas-cooled Reactors (AGR in the UK. Concrete specimens were conditioned at temperatures of 105 °C and 250 °C, to simulate the maximum thermal load expected to occur during a loss of coolant accident (LOCA. Following thermal treatment, these specimens along with an unconditioned control sample were characterised using micro-focus X-ray CT with a spatial resolution of 14.6 microns. The results indicate that the air void pore structure of the specimens experienced significant volume changes as a result of the increasing temperature. The increase in the porous volume was more prevalent at 250 °C. Alterations in air void size distributions were characterized with respect to the unconditioned control specimen. These findings appear to correlate with changes in the uni-axial compressive strength of the conditioned concrete.

  10. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

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

  11. Monitoring of prestressed concrete pressure vessels. 1. An overview of concrete embedment strain instrumentation and calibration test results for selected concrete embedment strain meters

    International Nuclear Information System (INIS)

    Naus, D.J.; Hurtt, C.C.

    1978-01-01

    The report presents results of calibration tests on strain meters. The approach was divided into two phases: (1) an overview of meter performance criteria for PCPV applications and techniques for strain measurements in concrete and (2) procurement of commercially available gages and their evaluation to assess the reliability of manufacturer-supplied calibration factors. Calibration test results for gages embedded in 15.2-cm-diam by 54-cm cylindrical concrete specimens indicated that calibration factors should be determined (verified) by embedding samples of the gages in test specimens fabricated using a representative mix and that further research should be conducted on other measurement techniques based on inductance, capacitance, semiconductors, and fluidic principles

  12. Dynamic analysis of crack growth and arrest in a pressure vessel subjected to thermal and pressure loading

    International Nuclear Information System (INIS)

    Brickstad, B.

    1984-01-01

    Predictions of crack arrest behaviour are performed for a cracked reactor pressure vessel under both thermal and pressure loading. The object is to compare static and dynamic calculations. The dynamic calculations are made using an explicit finite element technique where crack growth is simulated by gradual nodal release. Three different load cases and the effect of different velocity dependence on the crack propagation toughness are studied. It is found that for the analysed cases the static analysis is slightly conservative, thus justifying its use for these problems. (orig.)

  13. Non-linear analysis up to rupture of a model of a multi-cavity prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Rebora, B.; Uffer, F.; Zimmermann, T.

    1977-01-01

    Within the frame of a German-Swiss agreement concerning the project of a high-temperature nuclear plant (HHT), the Swiss Federal Institute for Reactor Research (EIR, in Wuerlingen) has developed an integrated variant of an helium-cooled high temperature reactor of 3x500 Mwe. A test on a model (1:20) of this prestressed concrete nuclear vessel with multiple cavities has been carried out under the supervision of 'Bonnard et Gardel ingenieurs-conseils SA (BG). The aim of this analysis is to determine the mechanism of ruin and ultimate load of the structure. In addition, comparison with the results of the test emphasizes the mathematical model with a view to its utilisation for the analysis of any prestressed concrete nuclear vessel. The principal interest of this paper is to show the accuracy of non-linear analysis of a complex massive structure with the test results and the evolution of the behaviour of the structure from the apparition of the first crack up to the ruin by rupture of the steel wires. (Auth.)

  14. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1978-01-01

    A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction

  15. Pressurized Vessel Slurry Pumping

    International Nuclear Information System (INIS)

    Pound, C.R.

    2001-01-01

    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

  16. Pressure vessel, in particular reactor pressure vessel

    International Nuclear Information System (INIS)

    1976-01-01

    The task to design the pre-stressing facility for a pressure vessel, especially for nuclear reactors, wiht pre-stressed jacket and pre-stressing facility, the latter one showing circumferential steel tendons supported polygonally on the outer side of the jacket by means of supporting shoes, in such a way that simply defined tensions can be generated and re-tensioning of the tendons can be carried out easily is solved according to the invention by keeping the circumference of the steel tendons fixed and by designing the supporting shoes as stressing shoes. The defined tensions are applied through the stressing shoes. (ORU) [de

  17. Sapphire tube pressure vessel

    Science.gov (United States)

    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.

  18. GOLD PRESSURE VESSEL SEAL

    Science.gov (United States)

    Smith, A.E.

    1963-11-26

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

  19. Pressure vessel integrity 1991

    International Nuclear Information System (INIS)

    Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

    1991-01-01

    This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

  20. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  1. Reactor pressure vessel nozzle

    Science.gov (United States)

    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.

  2. Attachment Fitting for Pressure Vessel

    Science.gov (United States)

    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.

  3. Reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  4. Support device in a pressure vessel, particularly a reactor pressure vessel, to support against horizontal forces

    International Nuclear Information System (INIS)

    Dorner, H.; Harand, E.; Scholz, M.; Scheler, R.

    1986-01-01

    In a support device on a reactor pressure vessel of a PWR, a guide leading downwards is provided in the vertical pellet cartridge on the floor side of the reactor pressure vessel, which is located in a fixed support pipe. The support pipe is supported on a concreted baseplate in the horizontal direction, and can be screwed to this. The support pipe has a plate-shaped support foot to support it on the baseplate. The reactor pressure vessel, normally resting on its main coolant duct in the plane normal to the axis, can be effectively supported against horizontal forces by this additional support, without preventing thermal movement in the longitudinal axis. (orig./HP) [de

  5. Special enclosure for a pressure vessel

    International Nuclear Information System (INIS)

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

    1993-01-01

    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

  6. Development of an integrated prestressed-concrete pressure vessel for water-cooled reactors (SBB type 'STERN' (star) with supporting boiler)

    International Nuclear Information System (INIS)

    Jueptner, G.; Kumpf, H.; Molz, G.; Neunert, B.; Seidl, O.

    1976-01-01

    This report goes into the reasons for selecting a 'STERN' (star) vessel configuration for accommodating a complete primary circuit including PWR, this involving the grouping of cylindrical pressure vessels of independent design into a star-shaped configuration with the central vessel housing the reactor core in the middle. This arrangement was made possible by application of the DYWIDAG-radial prestressing process generating controlled annular prestressing using existing presses and by an organic coupling of individual vessels. The liner, heat insulating and cooling system required for each vessel comprises a so-called support boiler, i.e. a hot liner not handicapped by the disadvantages of other systems. The support boiler is placed in the and PCV and has flat floor and cover surfaces. Temperature constraints are reduced to specific design requirements by means of radial gap permitting precise adjustment in conjunction with an axial expanding element comprising a multilayer diaphragm which is supported in operation. A detailed description is given of the PCPV, the support boiler and the cover used in the center vessel as well as of their design, the assembly and construction work is described and a summary presented of the quantities and estimated prices involved. Due to the absence of steam raising facilities adapted to meet the star-shaped configuration requirements, a study of satellite vessels was dispensed with, the design of which is in full accord with that of the center vessel. One part of the report is concerned with the calculation of the center vessel. (orig./HP) [de

  7. Residual Stress Estimation and Fatigue Life Prediction of an Autofrettaged Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyung Jin; Kim, Eun Kyum; Koh, Seung Kee [Kunsan Nat’l Univ., Kunsan (Korea, Republic of)

    2017-09-15

    Fatigue failure of an autofrettaged pressure vessel with a groove at the outside surface occurs owing to the fatigue crack initiation and propagation at the groove root. In order to predict the fatigue life of the autofrettaged pressure vessel, residual stresses in the autofrettaged pressure vessel were evaluated using the finite element method, and the fatigue properties of the pressure vessel steel were obtained from the fatigue tests. Fatigue life of a pressure vessel obtained through summation of the crack initiation and propagation lives was calculated to be 2,598 cycles for an 80% autofrettaged pressure vessel subjected to a pulsating internal pressure of 424 MPa.

  8. 46 CFR 119.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  9. 46 CFR 182.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  10. 46 CFR 169.249 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  11. Stress analysis of pressure vessels

    International Nuclear Information System (INIS)

    Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.

    1979-01-01

    This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)

  12. Pressure vessel and method therefor

    Science.gov (United States)

    Saunders, Timothy

    2017-09-05

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

  13. Pressure vessel having continuous sidewall

    Science.gov (United States)

    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. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Stoll, A.

    1976-01-01

    The invention relates to a pressure vessel which can be used for nuclear reactors and for chemical processing technologies. A grid of walls welded to each other, which is installed in the interior of the pressure vessel, is so attached to an outer jacket at several edges, that these edges exert a force on the wall of the vessel directed towards the interior. Only the out jacket resists the differential between the inner and outer pressures; the welded walls in the interior do not have to sustain any differential pressure. They create a larger number of inner spaces (or tubes) which can be individually accessible and each of which has a terminal element. (UWI) [de

  15. Level indicator for pressure vessels

    Science.gov (United States)

    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.

  16. Manufacturing technologies of PWR pressure vessels

    International Nuclear Information System (INIS)

    Qin Xubin

    1991-01-01

    Pressure vessels belong to the main component of PWR plants. Starting with describing the manufacture of pressure vessel components and their assembly, the manufacturing technologies of pressure vessels are briefly presented with regards to welding, heat treatment, inspections and testing. In addition, quality assurance during the manufacture is presented with emphasis

  17. Dismantling method for reactor pressure vessel and system therefor

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

    1994-01-01

    Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

  18. Reactor pressure vessel vented head

    Science.gov (United States)

    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.

  19. Innovations in prestressed concrete pressure vessel design

    International Nuclear Information System (INIS)

    Chow, P.Y.; Ngo, D.; Lin, T.Y.

    1979-01-01

    The study explored a new approach to the design of a high-pressure PCPV that accepts tension and tension cracks in the outer region of the PCPV. It examined the possibility of incorporating artificially-introduced preformed separations that pre-determined crack locations in the design as a method of controlling high tensile stresses generated by internal temperature and pressure. The results showed that the PCPV so designed was, in the extreme case of the DSV, approximately 70% cheaper than the 18 steel vessels of equivalent capacity it replaces. (orig.)

  20. Radiation embrittlement predictions for reactor pressure vessels

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Marston, T.U.

    1979-01-01

    The purpose of a surveillance program in a nuclear reactor pressure vessel is to provide an accurate measure of the changes in properties necessary to assess the structural integrity of the beltline region of the vessel as it is subjected to neutron irradiation damage. To understand problems surrounding a surveillance program, it is necessary to look at the concept of a structural integrity analysis. The structural integrity analysis can, of course, be only as accurate as the input data allows. Any errors in materials properties due to either errors in test technique to obtain these properties or to test specimens being exposed to environments different than actual operating condition will be reflected as errors in the actual structural integrity analysis. In a surveillance program, we must then be sure that we are measuring the proper material property to input to the fracture mechanics model as well as insure that the irradiation environment to which we subject the specimens will not cause an error in the prediction of the actual integrity margin of the pressure vessel beltline. The following text will question the appropriateness of using irradiated materials property data derived from high flux experiment, either in test reactors or advanced surveillance positions, to predict the actual material condition at the reactor vessel wall. Additionally, data will be presented showing several new correlation methods for deriving fracture toughness data from Charpy specimens, thus giving the proper input to the fracture mechanics model of a structural integrity analysis

  1. Support device in a pressure vessel, particularly a reactor pressure vessel, to support against horizontal forces. Abstuetzeinrichtung an einem Druckbehaelter, insbesondere einem Reaktordruckbehaelter, gegen Horizontalkraefte

    Energy Technology Data Exchange (ETDEWEB)

    Dorner, H.; Harand, E.; Scholz, M.; Scheler, R.

    1986-04-17

    In a support device on a reactor pressure vessel of a PWR, a guide leading downwards is provided in the vertical pellet cartridge on the floor side of the reactor pressure vessel, which is located in a fixed support pipe. The support pipe is supported on a concreted baseplate in the horizontal direction, and can be screwed to this. The support pipe has a plate-shaped support foot to support it on the baseplate. The reactor pressure vessel, normally resting on its main coolant duct in the plane normal to the axis, can be effectively supported against horizontal forces by this additional support, without preventing thermal movement in the longitudinal axis.

  2. Stabilizer for reactor pressure vessel

    International Nuclear Information System (INIS)

    Nakajima, Masataka.

    1991-01-01

    A plurality of flush springs and sleeves are disposed to a rod to be screw-coupled to a yoke disposed between gusset side walls. A vibration energy dissipation mechanism due to resiliently plastic deformation of metal is disposed between the gusset side wall and the sleeve of the rod. When earthquakes should occur and the distance between the gusset side wall and the sleeve is changed undergoing seismic input, the vibration energy is dissipated by the resilient plastic deformation of the vibration energy dissipation mechanism. The seismic response of the pressure vessel system is thus decreased to improve seismic resistivity. The reactor safety and reliability can be ensured. It is not necessary to strengthen supports for pipeline systems, and increase of the cost, including design and manufacture, can be avoided while giving no effects on the layout for the upper portion of reactor shielding walls. (N.H.)

  3. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  4. Spalling of concrete subjected to blast loading

    Directory of Open Access Journals (Sweden)

    Foglar M.

    2013-09-01

    Full Text Available This paper presents outcomes of the blast field tests of FRC and reinforced concrete specimens, which were performed in cooperation with the Czech Army corps and Police of the Czech Republic in the military training area Boletice. The numerical evaluation of the experiments focused on the spalling of concrete subjected to blast loading started after the first set of the tests, took almost 3 years and required further small-scale experiments performed in the labs of the Czech Technical University.

  5. Requirements on crack detection in pressurized vessels for ASME authorization

    International Nuclear Information System (INIS)

    Spremo, N.; Begovich, B.; Doko, A.

    1986-01-01

    The method is presented of training qualified personnel for nondestructive testing of pressure vessels. Personnel is divided according to experience and previous training into three groups of which each has its own educational programme. Written examinations in general knowledge and in specialized subjects and a practical examination in crack detection terminate the training. (E.S.). 1 fig., 4 tabs., 3 refs

  6. How to replace a reactor pressure vessel

    International Nuclear Information System (INIS)

    Huber, R.

    1996-01-01

    A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

  7. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  8. Results of reactor pressure vessels ISI

    International Nuclear Information System (INIS)

    Cepcek, S.

    1994-01-01

    To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

  9. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  10. USER SPECIFICATIONS FOR PRESSURE VESSELS AND TECHNICAL INTEGRITY

    Directory of Open Access Journals (Sweden)

    K.S. Johnston

    2012-01-01

    Full Text Available

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

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

  11. Motor-driven screwing and transporting tool for pressure vessels, especially pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    A screwing and transporting device for tensioning and loosening the reactor pressure vessel head is described. The advantage of the tool is its ability to unscrew the stud bolts from the lower part of the pressure vessel, too, and therefore make them accessible for in-service inspection. (TK) [de

  12. Curved and conformal high-pressure vessel

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    2010-10-01

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

  14. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1993-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described together with different aspects of the multielement integrity argument. The main revisions to the mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels are discussed. (author)

  15. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

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

  16. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

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

  17. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1992-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

  18. Internal Friction of Pressure Vessel Steel Embrittlement

    International Nuclear Information System (INIS)

    Van Ouytsel, K.

    2001-01-01

    The contribution consists of an abstract of a PhD thesis. The thesis contains a literature study, a description of the construction details of a new inverted torsion pendulum. This device was designed to investigate pressure-vessel steels at high amplitudes (10 -4 to 10 -2 ) and over a wide temperature range (90-700K) at approximately 1 Hz in the irradiated condition. Results of measurements on a variety of reactor pressure vessel steels by means of the torsion penduli are reported and interpreted

  19. Circular cylinders and pressure vessels stress analysis and design

    CERN Document Server

    Vullo, Vincenzo

    2014-01-01

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

  20. Application of dynamic relaxation and finite elements methods for the structural analysis of a scale model of a prestressed concrete pressure vessel; Aplicacao dos metodos de relaxacao dinamica e elementos finitos na analise estrutural de um modelo reduzido de vaso de pressao de concreto protendido

    Energy Technology Data Exchange (ETDEWEB)

    Tamura, Masaru

    1979-07-01

    A stress and strain analysis was made of a scale model of a Prestressed Concrete Pressure Vessel for a Boiling Water Reactor. The aim of this work was to obtain an experimental verification of the calculation method actually used at IPEN. The 1/10 scale model was built and tested at the Instituto Sperimentale Modelli e Structture, ISMES, Italy. The dynamic relaxation program PV2-A and the finite element programs , FEAST-1 have been used. A comparative analysis of the final results was made. A preliminary analysis was made for a simplified monocavity model now under development at IPEN with the object of confirming the data and the calculation method used. (author)

  1. Reactor pressure vessel with forged nozzles

    Science.gov (United States)

    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.

  2. 46 CFR 50.30-10 - Class I, I-L and II-L pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  3. Frost damage of concrete subject to confinement

    DEFF Research Database (Denmark)

    Hasholt, Marianne Tange

    2016-01-01

    When internal frost damage is observed in real concrete structures, the usual pattern is cracks with a preferred orientation parallel to the exposed surface. When exposing concrete with poor frost resistance to a standardised freeze/thaw test in the laboratory, the orientations of the resulting...

  4. Proactive life extension of pressure vessels

    Science.gov (United States)

    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

  5. Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

    International Nuclear Information System (INIS)

    Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

    1979-01-01

    A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

  6. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  7. Irradiation embrittlement of pressure vessel steels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Vacek, M.

    1975-01-01

    A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

  8. Analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A.N. de; Augusto, O.B.

    1985-01-01

    This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

  9. Composite Pressure Vessel Including Crack Arresting Barrier

    Science.gov (United States)

    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.

  10. Examination of VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Markulin, K.

    2008-01-01

    The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

  11. Pressure vessel integrity and weld inspection procedure

    International Nuclear Information System (INIS)

    Solomon, K.A.; Okrent, D.; Kastenberg, W.E.

    1975-01-01

    The primary objective of this paper is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an inter-relation between pressure vessel integrity, and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. A modified Markov process is employed and a computer code was written to obtain numerical results. The Markov process mathematically describes the following physical events. In a nuclear reactor pressure vessel weld, some defects may exist prior to the zeroth inspection (i.e., prior to vessel operation). During the zeroth inspection and repair processes, some of these defects are removed. During the first cycle of vessel operation, the existing defects may grow and some new defects may be generated. Those defects that are found at the first (and succeeding) inspection interval and warrant repair, are repaired. The above process continues through several operating cycles to the end of vessel life. During any inspection, only a portion of the welds may be inspected, and with less than perfect efficiency

  12. Code boiler and pressure vessel life assessment

    International Nuclear Information System (INIS)

    Farr, J.R.

    1992-01-01

    In the United States of America and in Canada, laws and controls for determining life assessment for continued operation of equipment exist only for those pressure vessels built to Section III and evaluated according to Section XI. In this presentation, some of those considerations which are made in the USA and Canada for deciding on life or condition assessment of boilers and pressure vessels designed and constructed to other sections of the ASME Boiler and Pressure Vessel Code are reviewed. Life assessment or condition assesssment is essential in determining what is necessary for continued operation. With no ASME rules being adopted by laws or regulations, other than OSHA in the USA and similar environmental controls in Canada, to control life assessment for continued operation, the equipment owner must decide if assessment is to be done and how much to do. Some of those considerations are reviewed along with methods and procedures to make an assessment along with a discussion of where the ASME B and PV Code currently stands regarding continued operation. (orig.)

  13. Preliminary study of an expert system for mechanical design of a pressure vessel

    International Nuclear Information System (INIS)

    Kasmuri, N.H.; Md Som, A.

    2006-01-01

    This paper describes a preliminary study of an expert system for mechanical design of a pressure vessel. The system supports the framework for the conceptual mechanical design from the initial stages within the design procedures. ASME Boiler and Pressure Vessel Code Section VIII Division 1 were applied as a design rule. The proposed methodology facilitates the development of knowledge base acquisition, knowledge base construction and the prototype implementation. This study characterizes a knowledge base (procedure) of mechanical design of a pressure vessel subjected to internal pressure including all design parameters; i.e. temperature, shell thickness, selection of materials of constructions, stress analysis procedure, support and ancillary items. The rationalization of the mechanical design is shown in the form of a schematic flow diagram. A Kappa PC expert system shell is used as a tool to develop the prototype software. It provides graphical representation for creating objects, hierarchies and rules for knowledge base used in pressure vessel design. (Author)

  14. 46 CFR 115.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... 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. [CGD 85-080, 61...

  15. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Science.gov (United States)

    2010-10-01

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

  16. 46 CFR 58.60-3 - Pressure vessel.

    Science.gov (United States)

    2010-10-01

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

  17. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  18. Compact insert design for cryogenic pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Dictionary of pressure vessel and piping technology

    International Nuclear Information System (INIS)

    Jentgen, L.; Schmitz, H.P.

    1986-01-01

    A specialised dictionary has been compiled containing the appropriate English and German terms in the following technical fields: materials science, welding, destructive and non-destructive testing, thermal and mass transfer, the design and construction in particular of pressure vessels, tanks, heat exchangers, piping, expansion joints, valves, and components associated with the above fields. This dictionary is the result of many years spent in evaluating technical terminology from the relevant American and British regulations, technical rules, standards, and specifications (see bibliography) and correlating these with the terminology of comparable German regulations, rules and standards, together with the essential technical literature. (orig.) [de

  20. Cylindrical pressure vessel constructed of several layers

    International Nuclear Information System (INIS)

    Yamauchi, Takeshi.

    1976-01-01

    For a cylindrical pressure vessel constructed of several layers whose jacket has at least one circumferential weld joining the individual layers, it is proposed to provide this at least at the first bending line turning point (counting from the weld between the jacket and vessel floor), which the sinusoidally shaped jacket has. The section of the jacket extending in between should be made as a full wall section. The proposal is based on calculations of the bending stiffness of cylindrical jackets, which could not yet be confirmed for jackets having several layers. (UWI) [de

  1. Reactor pressure vessel aging and countermeasures

    International Nuclear Information System (INIS)

    Leitz, C.

    1987-01-01

    The considerable aging effect on reactor pressure vessels is the effect of irradiation on material properties in the core beltline region. Modern LWRs in the Federal Republic of Germany are designed such that irradiation effects are very low. Countermeasures applicable separately or in combination for plants with higher than normally expected irradiation effects are described in three steps: first, examinations and calculations to extend the formal reactor lifetime by reducing over-conservative margins; second, changes in core design to reduce future irradiation effects; third, a procedure to recover irradiation effect on material properties already sustained. (orig./HP) [de

  2. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  3. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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

  4. Leak detector for reactor pressure vessel

    International Nuclear Information System (INIS)

    Morimoto, Mikio.

    1991-01-01

    A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

  5. Thermal ratcheting in pressure vessels and piping

    International Nuclear Information System (INIS)

    Kalnins, A.; Updike, D.P.

    1975-01-01

    During its lifetime, a nuclear power plant can experience cyclic thermal loading produced during start-ups and shut-downs. Under such conditions, pressure vessels, piping and other components can experience accumulative dimensional changes through thermal ratcheting. Such dimensional changes can occur when cyclic thermal loading is superposed upon steady mechanical loading. Simplified models based on one-dimensional plastic stress-strain relations, as proposed by Miller, Bree and Burgreen, are currently used by designers to estimate the likelihood and the amount of ratcheting during the life of the component. It is shown in this paper that the Miller-Bree-Burgreen one-dimensional model is applicable when the ratio of the steady membrane stresses is larger than one-half (Nsub(y)/Nsub(x)>1/2), which includes such significant applications as cylindrical and spherical vessels under internal pressure. For the stress ratios approaching minus one, the biaxial model predicts much higher ratchet strain. Examples where such states occur are the knuckle region in pressure vessels with torispherical heads and pipe connections under torsion. Designers should be alerted to such a limitation of the presently used ratcheting model. (Auth.)

  6. Pool critical assembly pressure vessel facility benchmark

    International Nuclear Information System (INIS)

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

    1997-07-01

    This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

  7. Feedwater control device for reactor pressure vessels

    International Nuclear Information System (INIS)

    Oonuma, Takeshi.

    1982-01-01

    Purpose: To prevent the generation of thermal stresses at the junction between a clean-up water pipe and a feedwater pipe. Constitution: Hot water containing impurities in a pressure vessel is caused to flow by a recycling pump through a heat exchanger, a cooler and a clean-up desalter and again by way of the heat exchanger into the feedwater pipe at the junction with the clean-up water pipe, where it is mixed with the feedwater passed by way of a feedwater heater and supplied to the pressure vessel. The feedwater temperature for the feedwater pipe and the set temperature for the clean-up water are compared with each other by using temperature sensors disposed to the feedwater pipe between the junction and the feedwater heater at the upstream of the junction. If the temperature difference is increased, for instance, upon transient state where the operation of the feedwater heater is not yet stabilized, the recycling pump is controlled to stop the supply of the clean-up water to the junction while flowing only the feedwater. This makes the temperature distribution uniform and prevents the generation of the thermal stresses at the junction, by which reactor safety can be improved. (Moriyama, K.)

  8. Nuclear reactor pressure vessel support system

    Science.gov (United States)

    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.

  9. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

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

    1994-01-01

    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 shallow 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 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 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

  10. Inspecting nuclear pressure vessels: the conundrum of minimizing risk

    International Nuclear Information System (INIS)

    Oestberg, G.

    1992-01-01

    The probability of a sudden, massive release of radioactivity from a light-water nuclear reactor through a breach of the containment is assessed on the basis of statistical data which partly consist of subjective estimates. This breach refers to the existence of crack-like defects remaining after a non-destructive examination of the main pressure vessel surrounding the reactor core. Two studies have recently been made of such sources of information about the effectiveness of non-destructive examination of pressure vessels with respect to defects. The results of these studies indicate that the data used as input in the probabilistic calculations do not possess the reliability that might be assumed from the assessments. This type of failure should therefore no longer be considered a de minimis case. In the present review the overconfidence in the efficiency of non-destructive examination is discussed from psychological, sociological and political science points of view. It is concluded that ingrained professional assumptions and values seem to be the main reason for the trust in the technology of inspection. However, there are also psychological constraints that can be understood only in their social and political contexts. (author)

  11. Seal analysis technology for reactor pressure vessel

    International Nuclear Information System (INIS)

    Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

    2009-01-01

    There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

  12. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    Biach, W.L.; Hill, R.; Hung, K.

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  13. A framework expert system for pressure vessels

    International Nuclear Information System (INIS)

    Wang, Y.C.; Qin, S.J.

    1989-01-01

    Expert systems, known as a powerful tool to those numerical problems accompanied with logical argumentation, are facing the era of extended application into the engineering fields beyond the classical scopes of diagnosis and consultation. With regard to pressure vessels design it seems that the most important task is to establish a general purpose frame based on a microcomputer skeleton system to meet the various requirements of different vessels. The authors have made an attempt to perform such a skeleton designated file, ESTOOL, in order to achieve the objectives of executing numerical calculation combined with logical reasoning, and attaining higher efficiency of rules searching process. It has been successfully patched to the design software package for jacketed vessel with stirring shaft. This paper presents the guiding concepts and basic structure of ESTOOL via knowledge acquisition subsystem and inference engine

  14. Computing the partial volume of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-15

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

  15. Recent experience reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    Vignes, A.

    1985-01-01

    This paper present the Framatome's recent experience in the manufacture of 1300 MWe PWR vessels; one shows how the very high standards of quality have been obtained to meet the stringent requirements. After a description of a pressure vessel, materials and forgings properties are presented. The nature and sequence of the main fabrication operations are reviewed. This paper deals after with the quality of welds, the preheating and post-heating equipment, the submerged arc welding process and procedures, the cladding process, and the under-clad cracking problems. Ultrasonic inspection procedures of the main welds are described with a comparison of RCCM (design and construction rules for mechanical components of PWR units) and Sizewell B specifications. Support of data on the reproductibility and effectiveness of ultrasonic examination and on the reliability given by repetitive inspection are presented

  16. Design of Saddle Support for Horizontal Pressure Vessel

    OpenAIRE

    Vinod Kumar; Navin Kumar; Surjit Angra; Prince Sharma

    2014-01-01

    This paper presents the design analysis of saddle support of a horizontal pressure vessel. Since saddle have the vital role to support the pressure vessel and to maintain its stability, it should be designed in such a way that it can afford the vessel load and internal pressure of the vessel due to liquid contained in the vessel. A model of horizontal pressure vessel and saddle support is created in ANSYS. Stresses are calculated using mathematical approach and ANSYS soft...

  17. Computations for the 1:5 model of the THTR pressure vessel compared with experimental results

    International Nuclear Information System (INIS)

    Stangenberg, F.

    1972-01-01

    In this report experimental results measured at the 1:5-model of the prestressed concrete pressure vessel of the THTR-nuclear power station Schmehausen in 1971, are compared with the results of axis-symmetrical computations. Linear-elastic computations were performed as well as approximate computations for overload pressures taking into consideration the influences of the load history (prestressing, temperature, creep) and the effects of the steel components. (orig.) [de

  18. Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

    International Nuclear Information System (INIS)

    1998-10-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

  19. Further fields of application for prestressed cast iron pressure vessels (PCIV)

    International Nuclear Information System (INIS)

    Guelicher, L.; Schilling, F.E.

    1977-01-01

    The redundancy of the prestressing system of prestressed structures as well as the clear separation of sealing and load-carrying functions of prestressed cast iron pressure vessels offer substantial advantages over conventional welded steel pressure vessels. Because of the temperature resistance of cast iron up to 400 0 C it is possible to build prestressed pressure vessels commercially as hot-working structures. The compressive strength of cast iron, which is 25 times as high as that of concrete allows for a very compact design of the PCIV. Further specific properties of the PCIV like pre-fabrication of the vessel in the production plant - made possible by a structure assembled from segments - short assembly periods at the construction site etc., may open more fields of application. - PCIV as pressurized storage tanks for the emergency shut down system in nuclear power stations. - PCIV as high pressure vessel for the chemical industry. - PCIV as energy storage. - PCIV for light water reactors. - PCIV as burst protection. It is concluded that the application of prestressed cast iron promises to be successful where either structures with large volumes and high pressures and/or temperatures are required or where aspects of safety allow for efficient use of prestressed structures. (Auth.)

  20. Temperature Response in Hardened Concrete Subjected to Tropical Rainforest Environment

    Directory of Open Access Journals (Sweden)

    E. I. Egba

    2017-06-01

    Full Text Available The objective of this paper is to characterize concrete micro-environment temperature response to the natural climate of the tropical rainforest. The peculiar warmth, high humidity, and low pressure nature of the tropical rainforest necessitated the present study. Temperature probes were inserted into concrete specimens subjected to the sheltered and unsheltered environment to measure the micro-environment temperature of the concrete, and study the hysteresis characteristics in relation to the climate temperature. Some mathematical relationships for forecasting the internal temperature of concrete in the tropical rainforest environment were proposed and tested. The proposed relationships were found reliable. It was observed that the micro-environment temperature was lower at the crest, and higher at the trough than the climate environment temperature with a temperature difference of 1-3 oC. Also, temperature response in concrete for the unsheltered micro-environment was 1.85 times faster than the response in the sheltered micro-environment. The findings of the study may be used to assist the durability assessment of concrete.

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

    Science.gov (United States)

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

    2014-01-01

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

  2. Reactor Pressure Vessel P-T Limit Curve Round Robin

    Energy Technology Data Exchange (ETDEWEB)

    Jang, C.H.; Moon, H.R.; Jeong, I.S. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This report is the summary of the analysis results for the P-T Limit Curve construction which have been subjected to the round robin analysis. The purpose of the round robin is to compare the procedure and method used in various organizations to construct P-T limit curve to prevent brittle fracture of reactor pressure vessel of nuclear power plants. Each Participant used its own approach to construct the P-T limit curve and submitted the results, By analyzing the results, the reference procedure for the P-T limit curve could be established. This report include the results of the comparison of the procedure and method used by the participants, and sensitivity study of the key parameters. (author) 23 refs, 88 figs, 17 tabs.

  3. Technical note: irradiation embrittlement of pressure vessel steels, analysis of the IAEA coordinated program results

    International Nuclear Information System (INIS)

    Hammad, F.H.; Ghoneim, M.M.; Abou-Zahra, A.

    1985-01-01

    The embrittlement of certain steels as the result of neutron irradiation has significance for evaluating the risks associated with pressurized thermal shock and other possible pressure vessel failure mechanisms, especially in the heat-affected zones of welds of older pressure vessels. A knowledge of the degree of embrittlement associated with a given integral fluence, usually expressed as an increase in the ductile-brittle transition temperature, is therefore needed to estimate the service life of pressure vessels subject to such embrittlement. This Technical Note describes a reanalysis of the results of the International Atomic Energy Agency coordinated program to measure this effect, which succeeded in explaining and reducing the very large degree of scatter in the results originally obtained in the measurements

  4. Composite Overwrapped Pressure Vessels, A Primer

    Science.gov (United States)

    McLaughlan, Pat B.; Forth, Scott C.; Grimes-Ledesma, Lorie R.

    2011-01-01

    Due to the extensive amount of detailed information that has been published on composite overwrapped pressure vessels (COPVs), this document has been written to serve as a primer for those who desire an elementary knowledge of COPVs and the factors affecting composite safety. In this application, the word "composite" simply refers to a matrix of continuous fibers contained within a resin and wrapped over a pressure barrier to form a vessel for gas or liquid containment. COPVs are currently used at NASA to contain high pressure fluids in propulsion, science experiments, and life support applications. They have a significant weight advantage over all metal vessels but require unique design, manufacturing, and test requirements. COPVs also involve a much more complex mechanical understanding due to the interplay between the composite overwrap and the inner liner. A metallic liner is typically used in a COPV as a fluid permeation barrier. The liner design concepts and requirements have been borrowed from all-metal vessels. However, application of metallic vessel design standards to a very thin liner is not straightforward. Different failure modes exist for COPVs than for all-metal vessels, and understanding of these failure modes is at a much more rudimentary level than for metal vessels.

  5. Radiation effects on reactor pressure vessel supports

    International Nuclear Information System (INIS)

    Johnson, 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

  6. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    Foerli, O.; Hernes, T.

    1978-01-01

    A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

  7. 46 CFR 176.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

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

  8. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  9. Development of automated welding process for field fabrication of thick walled pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, U A

    1981-01-01

    Research on automatic welding processes for the fabrication of thick-walled pressure vessels continued. A literature review on the subject was completed. A laboratory study of criteria for judging acceptable root parameters continued. Equipment for a demonstration facility to test the components and processes of the automated welding system has been specified and is being obtained. (LCL)

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

    Science.gov (United States)

    Kupiec, Chet; Espy, John

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

  11. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    Science.gov (United States)

    Ellis, Tayera

    2015-01-01

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

  12. Stress categorization in nozzle to pressure vessel connections finite elements models

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos de

    1999-01-01

    The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae for simple shell

  13. Chemical Safety Alert: Rupture Hazard of Pressure Vessels

    Science.gov (United States)

    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.

  14. Analysis of properties of WWER-440 reactor pressure vessel semiproducts

    International Nuclear Information System (INIS)

    Horacek, L.; Brumovsky, M.; Brynda, J.

    1988-01-01

    An analysis is made of selected tensile characteristics and of the chemical composition of steels 15Kh2MFA and 18Kh2MFA used for the manufacture of semifinished products of WWER-440 reactor pressure vessels with respect to the first 15 pressure vessels manufactured in Czechoslovakia. The results are analyzed with respect to the developmental trend (comparing the results of analyses of the first 5 and the first 15 pressure vessels manufactured by Skoda) and a different origin (comparing those by Skoda by the Izhorskij Zavod in the USSR). The effects of thickness of the semifinished product on tensile properties and of the chemical composition on radiation resitance of WWER-440 pressure vessels are considered in more detail. (author). 3 tabs., 3 refs

  15. High Toughness Light Weight Pressure Vessel, Phase I

    Data.gov (United States)

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

  16. Variability of mechanical properties of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Petrequin, P.; Soulat, P.

    1980-01-01

    Causes of variability of mechanical properties nuclear pressure vessel steels are reviewed and discussed. The effects of product shape and size, processing history and heat treatment are investigated. Some quantitative informations are given on the scatter of mechanical properties of typical pressure vessel components. The necessity of using recommended or standardized properties for comparing mechanical properties before and after irradiation in pin pointed. (orig.) [de

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

    International Nuclear Information System (INIS)

    Seifert, W.

    1980-01-01

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

  18. Common pressure vessel development for the nickel hydrogen technology

    Science.gov (United States)

    Holleck, G.

    1981-01-01

    The design of a pressure vessel nickel hydrogen cell is described. The cell has the following key features: it eliminates electrolyte bridging; provides for independent electrolyte management for each unit stack; provides for independent oxygen management for each unit stack; has good heat dissipation; has a mechanically sound and practical interconnection; and has the maximum in common with state of the art individual pressure vessel technology.

  19. Common-Pressure-Vessel Nickel-Hydrogen Battery Development

    OpenAIRE

    Otzinger, Burton; Wheeler, James

    1991-01-01

    The dual-cell, common-pressure vessel, nickel-hydrogen configuration has recently emerged as an option for small satellite nickel-hydrogen battery application. An important incentive is that the dual-cell, CPV configured battery presents a 30 percent reduction in volume and nearly 50 percent reduction in mounting footprint, when compared with an equivalent battery of individual pressure- vessel (IPV) cells. In addition energy density and cost benefits are significant. Eagle-Picher Industries ...

  20. Fragility assessment method of Concrete Wall Subjected to Impact Loading

    International Nuclear Information System (INIS)

    Hahm, Daegi; Shin, Sang Shup; Choi, In-Kil

    2014-01-01

    These studies have been aimed to verify and ensure the safety of the targeted walls and structures especially in the viewpoint of the deterministic approach. However, recently, the regulation and the assessment of the safety of the nuclear power plants (NPPs) against to an aircraft impact are strongly encouraged to adopt a probabilistic approach, i.e., the probabilistic risk assessment of an aircraft impact. In Korea, research to develop aircraft impact risk quantification technology was initiated in 2012 by Korea Atomic Energy Research Institute (KAERI). In this paper, for the one example of the probabilistic safety assessment approach, a method to estimate the failure probability and fragility of concrete wall subjected to impact loading caused by missiles or engine parts of aircrafts will be introduced. This method and the corresponding results will be used for the total technical roadmap and the procedure to assess the aircraft impact risk (Fig.1). A method and corresponding results of the estimation of the failure probability and fragility for a concrete wall subjected to impact loadings caused by missiles or engine parts of aircrafts was introduced. The detailed information of the target concrete wall in NPP, and the example aircraft engine model is considered safeguard information (SGI), and is not contained in this paper

  1. A structure for the protection of nuclear-reactor pressurized-vessels against rupture

    International Nuclear Information System (INIS)

    Marcellin, J.-P.; Aubert, Gilles

    1974-01-01

    Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

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

    International Nuclear Information System (INIS)

    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

  3. Updated embrittlement trend curve for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kirk, M.; Santos, C.; Eason, E.; Wright, J.; Odette, G.R.

    2003-01-01

    The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

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

    Science.gov (United States)

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

    2007-01-01

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

  5. Cooling device for upper lid of reactor pressure vessel

    International Nuclear Information System (INIS)

    Takayama, Kazuhiko.

    1997-01-01

    An upper lid of a reactor pressure vessel in a BWR type reactor has one or more annular cooling elements on the surface. The outer side thereof is covered by a temperature keeping frame. The cooling elements have an annular hollow shape, and cooling water is supplied to the hollow portion. As the cooling water supplied to the cooling elements, cooling water of a reactor auxiliary cooling system as a cooling water system incorporated in the reactor container or cooling water of a dehumidification system of the reactor container is used. A plurality of temperature sensors are disposed to various portions of the upper lid of the pressure vessel. A control device determines scattering of the temperature for the entire upper lid of the pressure vessel by temperature signals sent from the temperature sensors. Then, the amount of cooling water to be flown to each of the cooling elements is controlled so as to eliminate the scattering. (I.N.)

  6. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Science.gov (United States)

    2010-10-01

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

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

    Science.gov (United States)

    2010-10-01

    ... 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 as... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54...

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

    Directory of Open Access Journals (Sweden)

    Sedmak Aleksandar

    2016-01-01

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

  9. A prototype knowledge based system for pressure vessel design

    Energy Technology Data Exchange (ETDEWEB)

    Gunnarsson, L.

    1991-11-22

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD`s language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

  10. A prototype knowledge based system for pressure vessel design

    Energy Technology Data Exchange (ETDEWEB)

    Gunnarsson, L.

    1991-11-22

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

  11. Application of fracture mechanics to fatigue in pressure vessels

    International Nuclear Information System (INIS)

    Ghavami, K.

    1982-01-01

    The methods of application of fracture mechanics to predict fatigue crack propagation in welded structures and pressure vessels are described with the following objectives: i) To identify the effect of different variables such as crack tip plasticity, free surface, finite plate thickness, stress concentration and type of the structure, on the magnitude of stress intensity factor K in Welded joint. ii) To demonstrate the use of fracture mechanics for analysing fatigue crack propagation data. iii) To show how a law of fatigue crack propagation based on fracure mechanics, may be used to predict fatigue behavior of welded structures such as pressure vessel. (Author) [pt

  12. A prototype knowledge based system for pressure vessel design

    International Nuclear Information System (INIS)

    Gunnarsson, L.

    1991-01-01

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au)

  13. Conformable pressure vessel for high pressure gas storage

    Science.gov (United States)

    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.

  14. CONCRETE MIX DESIGN FOR STRUCTURES SUBJECTED TO EXPOSURE CLASS XC1 DEPENDING ON CONCRETE COVER

    Directory of Open Access Journals (Sweden)

    O. Yu. Cherniakevich

    2016-01-01

    Full Text Available The reinforced steel corrosion which is the most important problem of reinforced concrete structures durability is generally stipulated for carbonization of concrete surrounding it. Concrete cover calculation at the design stage is predicated one because of the differences in manufacturing conditions and use of constructions. The applying of the probabilistic approaches to the carbonation process modeling allows to get predicated grade of the depth of carbonization of concrete and, thus, to settle minimum concrete cover thickness for a given projected service life of a construction. The procedures for concrete mix design for different strength classes of concrete are described in the article. Current recommendations on assignment of concrete strength class as well as concrete cover are presented. The European Standard EN 206:2013 defines the content requirements for the concrete structures operated in the exposure class XC1, including the minimum values of water-cement ratio, minimum cement content, and minimum strength class of concrete. Since the standard does not include any basis or explanations of the requirements, we made an effort to develop a scientific justification for the mentioned requirements. We developed the probabilistic models for the process of carbonation of concrete based on the concrete mix which was designed using the software VTK-Korroziya. The reinforced concrete structures with concrete cover 20–35 mm operated in the most unfavorable conditions within the exposure class XC1 were analyzed. The corresponding probabilistic calculations of the depth of carbonated concrete are described in the article. 

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

    Science.gov (United States)

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

    2009-01-01

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

  16. Recent experiences and problems in conducting pressure vessel surveillance examinations

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1979-01-01

    Each of the commercial power reactors in the U.S.A. has a pressure vessel surveillance program. The purpose of the programs is to monitor the effects of radiation on the mechanical properties on the steel pressure vessels. A program for a given reactor includes a series of irradiation capsules containing neutron dosimeters and mechanical property specimens. The capsules are periodically removed during the life of the reactor and evaluated. The surveillance capsule examinations conducted to date have been valuable in assessing the effects of radiation on pressure vessels. However, a number of problems have been observed in the course of capsule examinations which potentially could reduce the maximum value of the data obtained. These problems are related to specimen design and preparation, capsule design and preparation, capsule installation and removal, capsule disassembly, specimen testing and evaluation, program documentation, and quality assurance. Examples of problems encountered in the preceding areas are presented in the present paper, and recommendations are made for minimization or prevention of these problems in future programs. Included in the recommendations is that appropriate ASTM standards, ASME Boiler and Pressure Vessel Code sections, and NRC regulations provide the appropriate framework for prevention of problems

  17. Design and Optimization of Filament Wound Composite Pressure Vessels

    NARCIS (Netherlands)

    Zu, L.

    2012-01-01

    One of the most important issues for the design of filament-wound pressure vessels reflects on the determination of the most efficient meridian profiles and related fiber architectures, leading to optimal structural performance. To better understand the design and optimization of filament-wound

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  19. Positron annihilation studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Brauer, G.; Liszkay, L.; Molnar, B.

    1988-01-01

    Several annealing studies by positron annihilation (Doppler broadening, lifetime) on neutron irradiated Cr-Mo-V reactor pressure vessel steels (Soviet type 15Kh2MFA) regarding the influences of irradiation temperature, fluence of fast neutrons as well as different impurity contents are presented and discussed. A possibility of explaining the positron annihilation data by irradiation induced carbide formation is proposed. (author)

  20. Some Observations on Damage Tolerance Analyses in Pressure Vessels

    Science.gov (United States)

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

    2017-01-01

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

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

    Science.gov (United States)

    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. Reliability aspects of radiation damage in reactor pressure vessel mterials

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1985-01-01

    The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

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

  6. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  7. Possible research program on a large scale nuclear pressure vessel

    International Nuclear Information System (INIS)

    1983-01-01

    The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    Wang, Jy-An John

    2010-01-01

    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.

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

    International Nuclear Information System (INIS)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

    1999-01-01

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

  11. Device for the burst protection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Daublebsky, P.

    1976-01-01

    The burst protection device has a hood over top and bottom of the pressure vessel with superimposed hinged supports lying in their turn against supporting rings which are connected with each other by vertical bracing. It is proposed to place an intermediate layer between hoods and vertical bracing absorbing thermal stresses, i.e. deforming plastically with gradually increasing pressure, but behaving like a rigid body in the case of shock loads. As a material lead e.g. is proposed. (UWI) [de

  12. The long-life design analysis of reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhang Ping

    2014-01-01

    Based on the fast neutron irradiation and thermal fatigue on reactor vessel ageing, the long-life design methods of reactor pressure vessel in EPR, AP1000, and CPR1000 are introduced. It is considered that adopting heavy reflector, increasing the downcomer of water gap, in-out refueling and increasing the toughness of the material are the methods to achieve the long-life running of RPV. (author)

  13. Why and how acoustic emission in pressure vessel first hydrotest

    International Nuclear Information System (INIS)

    Panzani, C.; Tonolini, F.; Villa, G.; Regis, V.

    1985-01-01

    The main advantages obtained performing the Acoustic Emission (AE) examination during pressure vessel first hydrotest are presented. The characteristics and performance of the AE instrumentation to be used for a correct test are illustrated. The main criteria for AE source characterization (location, typical AE parameters and their correlation with pressure value), the calibration and test procedures are discussed. The ndt post-test examinations and laboratory specimen experiments are also outlined. Personnel qualification requirements are finally indicated. (Author) [pt

  14. A classification system for pressure vessel shell failures

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1989-01-01

    A system for classifying failures of the shells of pressure vessels is presented. The classification system is based on the way a failure physically manifests itself and not on imputed economic or safety significance. It is believed the described way of classifying the failures is useful for transferring information from one situation to another. In assigning names to types of failure, the intention has been to adopt explicit definitions rather than supposed colloquial usage. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Advanced Approach of Reactor Pressure Vessel In-service Inspection

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Pajnic, M.

    2006-01-01

    The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

  17. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  18. Damage modelling in concrete subject to sulfate attack

    Directory of Open Access Journals (Sweden)

    N. Cefis

    2014-07-01

    Full Text Available In this paper, we consider the mechanical effect of the sulfate attack on concrete. The durability analysis of concrete structures in contact to external sulfate solutions requires the definition of a proper diffusion-reaction model, for the computation of the varying sulfate concentration and of the consequent ettringite formation, coupled to a mechanical model for the prediction of swelling and material degradation. In this work, we make use of a two-ions formulation of the reactive-diffusion problem and we propose a bi-phase chemo-elastic damage model aimed to simulate the mechanical response of concrete and apt to be used in structural analyses.

  19. Fiber Optic Acoustic Emission system for structural health monitoring of Composite Pressure Vessels, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Pressurized systems and pressure vessels used in NASA ground-based and flight-based applications including fuel tanks, composite overwrapped pressure vessels...

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

    Science.gov (United States)

    2010-10-01

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

  1. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

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

  2. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  3. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Science.gov (United States)

    2010-10-01

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

  4. 46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

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

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

    Science.gov (United States)

    2010-10-01

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

  6. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

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

  7. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 7 2010-07-01 2010-07-01 false Portable air receivers and other unfired pressure vessels... SHIPYARD EMPLOYMENT Portable, Unfired Pressure Vessels, Drums and Containers, Other Than Ship's Equipment § 1915.172 Portable air receivers and other unfired pressure vessels. (a) Portable, unfired pressure...

  8. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... operating plan must include: (i) A detailed description of the pressure vessel and all structures and... For those cases where materials are removed from the beltline of the pressure vessel, the stress...

  9. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Science.gov (United States)

    2010-10-01

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

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

    Czech Academy of Sciences Publication Activity Database

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

    2012-01-01

    Roč. 432, 1-3 (2012), s. 371-377 ISSN 0022-3115 R&D Projects: GA ČR GA101/09/1323 Institutional research plan: CEZ:AV0Z10100520 Keywords : neutron irradiation * steel degradation * nuclear reactor pressure vessel * magnetic NDT * magnetic minor hysteresis loops * Magnetic Adaptive Testing Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.211, year: 2012 http://dx.doi.org/10.1016/j.jnucmat.2012.09.006

  11. Behavior of Concrete Slab Track on Asphalt Trackbed Subjected to Thermal Load

    OpenAIRE

    Woo Young Jung; Seong Hyeok Lee; Jin Wook Lee; Bu Seog Ju

    2013-01-01

    Concrete track slab and asphalt trackbed are being introduced in Korea for providing good bearing capacity, durability to the track and comfortable rideness to passengers. Such a railway system has been designed by the train load so as to ensure stability. But there is lack of research and design for temperature changes which influence the behavior characteristics of concrete and asphalt. Therefore, in this study, the behavior characteristics of concrete track slab subjected to varying temper...

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

    Science.gov (United States)

    Green, Steven J.; Greene, Nathanael J.

    2012-01-01

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

  13. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  14. Corrosion fatigue of pressure vessel steel A533B. Part

    International Nuclear Information System (INIS)

    Gott, K.

    1978-08-01

    A survey is presented of the current work, both reported and in progress, to study the corrosion fatigue of the pressure vessel steel A533B C1 1 in a BWR environment. This report is based on the available literature, laboratory visits, and attendance at the Metals Society conference (April 1978). The parameters which are considered to have the most influence on the fatigue crack growth rate are considered in turn. Because of the sparsity of the available information an International Cooperative Group on Cyclic Crack Growth Rate Testing and Evaluation has been established. The initial work of the group and Swedish participation are described. (author)

  15. Prevention of catastrophic failure in pressure vessels and pipings

    International Nuclear Information System (INIS)

    Rintamaa, R.; Wallin, K.; Ikonen, K.; Toerroenen, K.; Talja, H.; Keinaenen, H.; Saarenheimo, A.; Nilsson, F.; Sarkimo, M.; Waestberg, S.; Debel, C.

    1989-01-01

    The fracture resistance and integrity of pressure-loaded components have been assessed in a Nordic research programme. Experiments were performed to validate the computational fracture assessment analysis. Two tests were also conducted on a large decommissioned pressure vessel from an oil refinery plant. Different fracture assessment methods were developed and subsequently applied to the tested components. Interlaboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finit element calculations and fracture mechanics testing. The transferability of material parameters derived from small specimens with simple crack geometries to more realistic crack geometries in real components has been verified. (author)

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

    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.

  18. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Science.gov (United States)

    Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

    2018-04-01

    Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

  19. Experimental on moisture migration and pore pressure formation of concrete members subjected to high temperature

    International Nuclear Information System (INIS)

    Nagao, Kakuhiro; Nakane, Sunao

    1993-01-01

    The experimental studies concerning temperature, moisture migration, and pore pressure of mass concrete mock-up specimens heated up to high temperature at 110degC to 600degC, were performed, so as to correctly estimate the moisture migration behaviour of concrete members subjected to high temperature, which is considered significantly influenced on physical properties of concrete. As a results, it is confirmed that the moisture migration behavior of concrete members can be explained by temperature and pore pressure, and indicate the characteristics both sealed condition (dissipation of moisture is prevented) and unsealed condition (dissipation of moisture occur). (author)

  20. Analysis of Concrete Slabs Subject to Concentrated Loads | Albrecht ...

    African Journals Online (AJOL)

    A simplified procedure for the calculation of the additional reinforcement which is required under concentrated loads acting on concrete slabs is provided. The introduced method can also be used for the treatment of line loads. Conditions for the application of this procedure are stated and tables for necessary design values ...

  1. Chloride Transport in OPC Concrete Subjected to the Freeze and Thaw Damage

    Directory of Open Access Journals (Sweden)

    Ki Yong Ann

    2017-01-01

    Full Text Available To predict the durability of a concrete structure under the coupling degradation consisting of the frosting and chloride attack, microstructural analysis of the concrete pore structure should be accompanied. In this study, the correlation between the pore structure and chloride migration for OPC concrete was evaluated at the different cement content in the concrete mix accounting for 300, 350, and 400 kg/m3 at 0.45 of a free water cement ratio. The influence of frosting damage on the rate of chloride transport was assessed by testing with concrete specimens subjected to a rapid freezing and thawing cyclic environment. As a result, it was found that chloride transport was accelerated by frost damage, which was more influential at the lower cement content. The microscopic examination of the pore structure showed that the freezing environment increased the volume of the large capillary pore in the concrete matrix.

  2. Microseismic Signal Characterization and Numerical Simulation of Concrete Beam Subjected to Three-Point Bending Fracture

    Directory of Open Access Journals (Sweden)

    Nuwen Xu

    2015-01-01

    Full Text Available To study the generation mechanism and failure mode of cracks in mass concrete, microseismic monitoring is conducted on the fracture processes of the three-point bending roller compacted concrete (RCC beam of Guanyinyan hydropower station. The spectrum characteristics of microseismic signals in different deformation and failure stages of the concrete beam are analyzed, and the identification method of the fracture stages and crack propagation precursors of concrete beam is established. Meanwhile, the Realistic Failure Process Analysis code (RFPA is adopted to simulate and analyze the entire failure processes of concrete beam from its cracks initiation, development, propagation, and coalescence, until macroscopic fractures formation subjected to three-point bending test. The relation curve of the load, loaded displacement, and acoustic emission (AE of concrete beam in the three-point bending test is also obtained. It is found that the failure characteristics of concrete beam obtained from numerical experiments agree well with the field physical test results. The heterogeneity of concrete is the major cause of zigzag propagation paths of beam cracks subjected to three-point bending tests. The results lay foundation for further exploring the formation mechanism of dam concrete cracks of Guanyinyan hydropower station.

  3. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

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

    Science.gov (United States)

    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%).

  5. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  6. Interpretation of Strain Measurements on Nuclear Pressure Vessels

    DEFF Research Database (Denmark)

    Andersen, Svend Ib Smidt; Engbæk, Preben

    1980-01-01

    Selected results from strain measurements on four nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts. The resu......Selected results from strain measurements on four nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts....... The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as a detailed knowledge of the behaviour of the signal...... from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour...

  7. Strain ageing in welds of nuclear pressure vessels

    International Nuclear Information System (INIS)

    Otterberg, R.; Karlsson, C.

    1979-01-01

    Static and dynamic strain ageing have been investigated on submerged-arc welds and repair welds from plates of the pressure vessel steel A 533B. The results permit the determination of the worst strain ageing conditions existing in a nuclear pressure vessel. Static strain ageing was investigated by means of data from tension tests, hardness measurements and Charpy-V impact properties for prestrained and aged material for ageing temperatures from room temperature to 350 deg C and ageing times up to 1000h. Dynamic strain ageing was investigated by tensile tests up to 350 deg C at different strain rates. At the most static strain ageing was found to increase the impact transition temperature from -75 deg C in the as-received condition to -55 deg C after prestraining and ageing for the plate material, from -35 to -10 deg C for the submerged arc weld and from -90 to -40 deg C for the repair weld. Approximately 10 deg C of the deleterious effect is due to the effect of ageing for the two former materials whereas the corresponding figure for the repair weld amounts to 35 deg C. The dynamic strain ageing is strongest at very low strain rates at temperatures just below 300 deg C. The effect of strain ageing can be reduced by stress relief heat treatment or by other means decreasing the content of nitrogen in solution. (author)

  8. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  9. Treating asphericity in fuel particle pressure vessel modeling

    Science.gov (United States)

    Miller, Gregory K.; Wadsworth, Derek C.

    1994-07-01

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

  10. Calculation of reactor pressure vessel fluence using TORT code

    International Nuclear Information System (INIS)

    Shin, Chul Ho; Kim, Jong Kyung

    1998-01-01

    TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Unit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) for all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library, BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The maximum fast neutron fluence calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effective full power days is 1.784x10 18 n/cm 2 . The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60 cm below the midplane at zero degree

  11. Scale modeling of reinforced concrete structures subjected to seismic loading

    International Nuclear Information System (INIS)

    Dove, R.C.

    1983-01-01

    Reinforced concrete, Category I structures are so large that the possibility of seismicly testing the prototype structures under controlled conditions is essentially nonexistent. However, experimental data, from which important structural properties can be determined and existing and new methods of seismic analysis benchmarked, are badly needed. As a result, seismic experiments on scaled models are of considerable interest. In this paper, the scaling laws are developed in some detail so that assumptions and choices based on judgement can be clearly recognized and their effects discussed. The scaling laws developed are then used to design a reinforced concrete model of a Category I structure. Finally, how scaling is effected by various types of damping (viscous, structural, and Coulomb) is discussed

  12. Analysis of reinforced concrete structures subjected to aircraft impact loading

    International Nuclear Information System (INIS)

    Bauer, J.; Scharpf, F.; Schwarz, R.

    1983-01-01

    Concerning the evaluation of the effects of aircraft impact loading on the reactor building and the contained equipment special interest belongs to both the characteristic of loading conditions and the consideration of the nonlinear behaviour of the local impacted area as well as the overall behaviour of the structure. To cover this extensive scope of problems the fully 3-dimensional code DYSMAS/L was prepared for the analysis of highly dynamic continuum mechanics problems. For this totally Lagrangian description, derived and tested in the field of the simulation of impact phenomena and penetration of armoured structures, an extension was made for the reasonable modelling of the material behaviour of reinforced concrete. Conforming the available experimental data a nonlinear stress-strain curve is given and a continuous triaxial failure-surface is composed which allows cracking of concrete in the tensile region and its crushing in the compressive mode. For the separately modeled reinforcement an elastic-plastic stress-strain relationship with kinematic hardening is used. (orig./RW)

  13. Stressing arrangement for pressure vessels with prestressed jacket

    International Nuclear Information System (INIS)

    1976-01-01

    Previously used stressing arrangements for pressure vessels assembled from segments are improved by producing defined prestressing forces. The manufactured elements made, for example, of cast iron, which make up the containment wall, carry one (or more) of the prestressing arrangements described in detail on the outside. The stressing arrangements of one position of the wall elements lie in the same horizontal plane around the container. A stressing cable or rope without or with only small prestressing is positioned around the prestressing arrangements lying at the same height. By operating the stressing piston arrangement of each device a required pressure is exerted on each wall element, and is permanently fixed by tightening self locking wedges or screws. (ORU) [de

  14. Ductile fracture estimation of reactor pressure vessel under thermal shock

    International Nuclear Information System (INIS)

    Takahashi, Jun; Sakai, Shinsuke; Okamura, Hiroyuki

    1990-01-01

    This paper presents a new scheme for the estimation of unstable ductile fracture of a reactor pressure vessel under thermal shock conditions. First, it is shown that the bending moment applied to the cracked section can be evaluated by considering the plastic deformation of the cracked section and the thermal deformation of the shell. As the contribution of the local thermal stress to the J-value is negligible, the J-value under thermal shock can be easily evaluated by using fully plastic solutions for the cracked part. Next, the phenomena of ductile fracture under thermal shock are expressed on the load-versus-displacement diagram which enables us to grasp the transient phenomena visually. In addition, several parametrical surveys are performed on the above diagram concerning the variation of (1) thermal shock conditions, (2) initial crack length, and (3) J-resistance curve (i.e. embrittlement by neutron irradiation). (author)

  15. Marine reactor pressure vessels dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Mount, M.E.

    1997-01-01

    Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  17. Fracture probability evaluation of a LWR pressure vessel

    International Nuclear Information System (INIS)

    Grandemange, J.; Pellissier-Tanon, A.; Quero, J.; Carnino, A.; Dufresne, J.

    1978-01-01

    Fracture probability evaluation, of a LWR pressure vessel have been performed in the past, using statistical data from conventional plant. A more accurate evaluation has been requested in 1976 from the SCSIN to the CEA. With this object, a joint collaboration agreement has been signed between CEA, EURATOM/ISPRA and FRAMATOME. The whole program proceeding from this agreement is managed by a joint board including the three partners. The basic objective of this program is to develop a method which integrates, or makes it possible to integrate at a later stage, the greatest number of significant parameters. Also, in order to prepare the practical applications, a special effort is being made to collect the data corresponding to these parameters. Parallel basic research program have been launched in order to clarify our knowledge on some important parts of the main factors contributing to the evaluation. The results of this research will be progressively introduced into the method or will help checking its validity

  18. Supplementary surveillance programme for WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Novosad, P.

    1994-01-01

    Paper gives a short analysis of a standard surveillance program for WWER440/213 type reactor pressure vessels. The program suggests the surveillance specimens are located in the region of high lead factor and thus they are suitable for surveillance for only up to about 5 years. For better and effective use of the results, their re-evaluation, based on gamma scanning of irradiated specimens as well as on testing of reconstituted irradiated specimens with the goal to determine the static and dynamic fracture toughness has been opened. New, supplementary surveillance program for the vessels operated in Czech Republic was designed: archive pieces of specimens (for further reconstitution after irradiation) will be irradiated in sites with low lead factor and then tested to determine the standard (impact) and fracture toughnesses

  19. Needs for evaluated covariance data for reactor pressure vessel dosimetry

    International Nuclear Information System (INIS)

    Maerker, R.E.; Broadhead, B.L.; Wagschal, J.J.

    1992-01-01

    This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

  20. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    Arutyunjan, R.; Kabalevsky, S.; Kiselev, V.; Serov, A.

    1997-01-01

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

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

    Science.gov (United States)

    Prosser, William H.

    2014-01-01

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

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

    Science.gov (United States)

    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.

  3. Stable and unstable crack growth in pressure vessel models

    International Nuclear Information System (INIS)

    Smith, G.C.; Canonico, D.A.; Stelzman, W.J.

    1978-01-01

    Three identical steel pressure vessels with 254-mm (10-in.) dia, 38-mm (1.5-in.) wall thicknesses and long, deep machined and sharpened axially oriented flaws were tested at three different temperatures. The vessels were assembled by electron-beam welding cylindrical sections with substantially different toughnesses due to different heat treatments. Crack extension initiated in relatively brittle sections, and the cracks extended both stably and unstably, depending on test temperature, toward the tougher sections where crack arrest did and did not occur. Charpy impact specimens and both slow-bend and dynamic precracked Charpy specimens were used for material characterization. The behavior of the vessels is described and related to the Charpy data

  4. Preparation of the Shippingport reactor pressure vessel shipping package

    International Nuclear Information System (INIS)

    Yannitell, D.M.

    1988-01-01

    Shippingport Station Decommissioning Project is the removal and shipment the Reactor Pressure Vessel (PRV) and its associated Neutron Shield Tank (NST) to the government owned Hanford Reservation in Richland, Washington. Engineering studies considered the alternatives for removal and shipment of the RPV/NST. These included segmentation for subsequent truck shipments, and one-piece removal with barge or rail shipment. Although the analysis indicated that current technology could be utilized to accomplish either alternative, one-piece removal of the RPV was selected as the safest, most cost effective method. When compared to segmentation, it was estimated that one-piece removal would reduce the duration of the Project by 1 year, reduce cost by $4 M, and result in a savings of radiation exposure of 150 man-Rem. Rail transportation of an integral RPV/NST package is not feasible due to the physical size of the package. 5 refs., 1 fig

  5. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  6. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    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

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

    Science.gov (United States)

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

    1973-01-01

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

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

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

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

  9. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

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

    2009-01-01

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

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

    Science.gov (United States)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, V., E-mail: Vladimir.Slugen@elf.stuba.sk [Institute of Nuclear and Physical Engineering, Slovak University of Technology, Ilkovičova 3, 81219 Bratislava (Slovakia); Hein, H. [AREVA NP GmbH, Paul Gossen Strasse 100, 91 001 Erlangen (Germany); Sojak, S.; Simeg Veterníková, J.; Petriska, M.; Sabelová, V.; Pavúk, M.; Hinca, R.; Stacho, M. [Institute of Nuclear and Physical Engineering, Slovak University of Technology, Ilkovičova 3, 81219 Bratislava (Slovakia)

    2013-11-15

    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. PAS techniques can be effectively applied for evaluation of microstructural changes caused by extreme external loads (characterized by high dpa values) by proton implantation, with aim to simulate irradiation and for the evaluation of the effectiveness of post-irradiation thermal treatments. We used our actual and previous results, collected during last 20 years from measurements of different RPV-steels in “as received”, irradiated and post-irradiation annealed state and compare them with the aim to contribute to general knowledge based on experimental PAS data. Actual results from irradiated German and Russian steels confirmed that no large voids or vacancy clusters were formed at defined irradiation conditions stated according to the real operational conditions at nuclear power plants. This indicate the fact that vacancy type defects bear hardly any responsibility for radiation-induced hardening and embrittlement of reactor pressure vessel steels and does not affect significantly the long-term operation of nuclear power plants from safety point of view.

  13. Reliability of Arch dams subject to concrete swelling

    Energy Technology Data Exchange (ETDEWEB)

    Ramos, J.M.; Silva, H.S.; Pinho, S. de [Laboratorio Nacional de Engenharia Civil (LNEC), Lisboa (Portugal)] [and others

    1995-12-31

    In this report, results of several studies are presented. The main aim of those studies was to assess the reliability of the three arch dams, in which swelling occurred due to alkali- aggregate reactions in various stages of development and having different effects on their reliability: the Cahora-Bassa dam, in Mozambique, where swelling accumulated up to the moment are very moderate and their development is apparently homogeneous; Santa-Luzia dam, in Portugal, where accumulated swelling have already considerable magnitude, nevertheless, important fissuration has not been observed up to the moment due to the homogeneous development of the swelling process; Alto-Ceira dam, also in Portugal, where accumulated swelling have also considerable magnitude but with a heterogeneous development, causing in conjunction with thermal variations important fissuration. Mention is made of mineralogical, chemical and petrographic analyses carried out for identification of the nature of reactions developed in each case and the back-analysis and other technics used in the assessment of the magnitude and distribution of swelling. Results are presented of measurement tests of the ultrasonic pulse velocity, used both in the assessment of alterations in the physical properties of concretes and in the determination of the depth of fissuration. Results are also presented of tests for characterisation of the rheology of integral concrete. Lastly, considerations are made about the reliability of the works on the basis of studies and the results of analyses of the state of stress, performed by means of the finite element method, by assuming for either visco-elastic or visco-elastic-plastic behaviour.

  14. NDE and Stress Monitoring on Composite Overwrapped Pressure Vessels, Phase I

    Data.gov (United States)

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

  15. Nonlinear analysis of reinforced concrete structures subjected to high temperature and external load

    International Nuclear Information System (INIS)

    Sugawara, Y.; Goto, M.; Saito, K.; Suzuki, N.; Muto, A.; Ueda, M.

    1993-01-01

    A quarter of a century has passed since the finite element method was first applied to nonlinear problems concerning reinforced concrete structures, and the reliability of the analysis at ordinary temperature has been enhanced accordingly. By contrast, few studies have tried to deal with the nonlinear behavior of reinforced concrete structures subjected to high temperature and external loads simultaneously. It is generally known that the mechanical properties of concrete and steel are affected greatly by temperature. Therefore, in order to analyze the nonlinear behavior of reinforced concrete subjected to external loads at high temperature, it is necessary to construct constitutive models of the materials reflecting the influence of temperature. In this study, constitutive models of concrete and reinforcement that can express decreases in strength and stiffness at high temperature have been developed. A two-dimensional nonlinear finite element analysis program has been developed by use of these material models. The behavior of reinforced concrete beams subjected simultaneously to high temperature and shear forces were simulated using the developed analytical method. The results of the simulation agreed well with the experimental results, evidencing the validity of the developed material models and the finite element analysis program

  16. Creep of A508/533 Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Richard Wright

    2014-08-01

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

  17. Performance of Hybrid Steel Fibers Reinforced Concrete Subjected to Air Blast Loading

    Directory of Open Access Journals (Sweden)

    Mohammed Alias Yusof

    2013-01-01

    Full Text Available This paper presents the results of the experimental data and simulation on the performance of hybrid steel fiber reinforced concrete (HSFRC and also normal reinforced concrete (NRC subjected to air blast loading. HSFRC concrete mix consists of a combination of 70% long steel hook end fibre and also 30% of short steel hook end fibre with a volume fraction of 1.5% mix. A total of six concrete panels were subjected to air blast using plastic explosive (PE4 weighing 1 kg each at standoff distance of 0.3 meter. The parameters measured are mode of failure under static and blast loading and also peak overpressure that resulted from detonation using high speed data acquisition system. In addition to this simulation work using AUTODYN was carried out and validated using experimental data. The experimental results indicate that hybrid steel fiber reinforced concrete panel (HSFRC possesses excellent resistance to air blast loading as compared to normal reinforced concrete (NRC panel. The simulation results were also found to be close with experimental data. Therefore the results have been validated using experimental data.

  18. Assessing and updating the reliability of concrete bridges subjected to spatial deterioration - principles and software implementation

    DEFF Research Database (Denmark)

    Schneider, Ronald; Fischer, Johannes; Bügler, Maximilian

    2015-01-01

    Inspection and maintenance of concrete bridges is a major cost factor in transportation infrastructure, and there is significant potential for using information gained during inspection to update predictive models of the performance and reliability of such structures. In this context, this paper ...... is a step towards developing a software tool that can be used by engineering practitioners to perform reliability assessments of ageing concrete bridges and update their reliability with inspection and monitoring data.......Inspection and maintenance of concrete bridges is a major cost factor in transportation infrastructure, and there is significant potential for using information gained during inspection to update predictive models of the performance and reliability of such structures. In this context, this paper...... presents an approach for assessing and updating the reliability of prestressed concrete bridges subjected to chloride-induced reinforcement corrosion. The system deterioration state is determined based on a Dynamic Bayesian Network (DBN) model that considers the spatial variability of the corrosion process...

  19. Moessbauer effect on a reactor pressure vessel irradiated with fast neutrons

    International Nuclear Information System (INIS)

    Yi, Y. G.; Kim, H. S.; Hong, C. Y.; Jang, K. S.; Yoo, Y. B.

    1998-01-01

    To estimate the integrity of a nuclear-reactor pressure vessel (RPV) subjected to neutron irradiation, we studied the changes in the material properties of RPV steel irradiated by fast neutrons (E n > 1 MeV) at a level of 0 ∼ 10 18 n/cm 2 . An X-ray diffraction experiment was done to investigate the changes in the phases of all specimens, nd the influence of the neutron irradiation on the RPV steel studied by the Mossbauer effect. The absorption area decreased rapidly for doses above 10 17 n/cm 2 , but at site 3 the value of the isomer shift and the quadrupole splitting increased. The magnetic hyperfine field at site 1 and 2 varied only within ±3 %

  20. ANSI/AIAA S-081A, Pressure Vessel Standards Implementation Guidelines

    Science.gov (United States)

    Greene, Nathanael J.

    2009-01-01

    The stress rupture specification for Composite Overwrapped Pressure Vessels (COPV) is discussed. The composite shell of the COPV shall be designed to meet the design life considering the time it is under sustained load. A Mechcanical Damage Control Plan (MDCP) shall be created and implemented that assures the COPV will not fail due to mechanical damage due to manufacturing, testing, shipping, installation, or flight. Proven processes and procedures for fabrication and repair shall be used to preclude damage or material degradation during material processing, manufacturing operations, and refurbushment.Selected NDI techniques for the liner and/or boss(es) shall be performed before overwrapping with composite. When visual inspection reveals mechanical damage or defects exceeding manufacturing specification levels (and standard repair procedures), the damaged COPV shall be submitted to a material review board (MRB) for disposition. Every COPV shall be subjected to visual and other non-destructive inspection (NDI), per the inspection plan.

  1. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  2. Ver reactor pressure vessel neutron exposure evaluation and monitoring

    International Nuclear Information System (INIS)

    Osmera, B.

    1992-01-01

    The surveillance neutron exposure monitoring system In VVER-440, V 213 series, consists of 54 Fe(n,p), 63 Cu(n,α and 93 Nb(n,n) detectors. The Fe detector is loaded for one year Cone fuel cycle) Irradiation only. The evaluation of the neutron monitor results including transformation to the critical locations of the reactor pressure vessel (RPV) can be based on the direct estimations of the neutron spectra and led factors (and other important space energy indices) in the mock-up experiments carried out. in the LR-0 experimental reactor, N.R.I. Rez. Because of high acceleration factor (about 11 for neutrons with energy above 0.5 MeV) the specimens are pulled out within five years of VVER-440 operation. After that the ex-vessel monitoring should be applied for RPV exposure determination. A recommended procedure (guide) for the RPV neutron exposure evaluation has been completed recently. All previous experimental data (the surveillance and ex-vessel monitors, I.R-0 mock-up measurements) have been revised arid reevaluated using the IRDF-90 (IAEA) cross section data. A brief review of this work is presented in the paper. Some comments relating to the RPV dosimetry of VVER-440, series 230, and VVER-1000 reactors are also done. (author)

  3. Underwater television camera for monitoring inner side of pressure vessel

    International Nuclear Information System (INIS)

    Takayama, Kazuhiko.

    1997-01-01

    An underwater television support device equipped with a rotatable and vertically movable underwater television camera and an underwater television camera controlling device for monitoring images of the inside of the reactor core photographed by the underwater television camera to control the position of the underwater television camera and the underwater light are disposed on an upper lattice plate of a reactor pressure vessel. Both of them are electrically connected with each other by way of a cable to rapidly observe the inside of the reactor core by the underwater television camera. The reproducibility is extremely satisfactory by efficiently concentrating the position of the camera and image information upon inspection and observation. As a result, the steps for periodical inspection can be reduced to shorten the days for the periodical inspection. Since there is no requirement to withdraw fuel assemblies over a wide reactor core region, and the device can be used with the fuel assemblies being left as they are in the reactor, it is suitable for inspection of detectors for nuclear instrumentation. (N.H.)

  4. Shallow-crack toughness results for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Theiss, T.J.; Shum, D.K.M.; Rolfe, S.T.

    1992-01-01

    The Heavy Section Steel Technology Program (HSST) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. To complete this investigation, techniques were developed to determine the fracture toughness from shallow-crack specimens. A total of 38 deep and shallow-crack tests have been performed on beam specimens about 100 mm deep loaded in 3-point bending. Two crack depths (a ∼ 50 and 9 mm) and three beam thicknesses (B ∼ 50, 100, and 150 mm) have been tested. Techniques were developed to estimate the toughness in terms of both the J-integral and crack-tip opening displacement (CTOD). Analytical J-integral results were consistent with experimental J-integral results, confirming the validity of the J-estimation schemes used and the effect of flaw depth on fracture toughness. Test results indicate a significant increase in the fracture toughness associated with the shallow flaw specimens in the lower transition region compared to the deep-crack fracture toughness. There is, however, little or no difference in toughness on the lower shelf where linear-elastic conditions exist for specimens with either deep or shallow flaws. The increase in shallow-flaw toughness compared with deep-flaw results appears to be well characterized by a temperature shift of 35 degree C

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    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. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1998-01-01

    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 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) 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 a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

  7. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  8. Radiation-induced embrittlement in light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.; McElroy, W.N.

    1987-01-01

    In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integry is a significant economic consideration because the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant. In addition to plant life considerations, LWR-PV embrittlement creates significant cycle-to-cycle impact through the restriction of normal heat-up and cool-down reactor operations. Recent LWR-PV benchmark experiments are analyzed. On this bases, it is established that an exponential representation accurately describes the spatial dependence of neutron exposure in LWR-PV. Implications produced by simple exponental behavior are explained and trend-curve models for the predictions of PV embrittelment are derived. These derivations provide for a clearer understanding and assessment of the assumptions underlying these trend-curve models. It is demonstrated that LWR-PV embrittlement possesses significant material dependence. (orig.)

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

    Science.gov (United States)

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

    2009-01-01

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

  10. Break location influence in pressure vessel SBLOCA scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Querol, Andrea; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

    2017-07-01

    The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused in simulating different hypothetical accidental scenarios in the PV using the thermalhydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios. (author)

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

    International Nuclear Information System (INIS)

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

    1996-06-01

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

  12. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

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

    International Nuclear Information System (INIS)

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

    1994-08-01

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

  14. Ultrasonic examination of austenitic welds at reactor pressure vessels

    International Nuclear Information System (INIS)

    Edelmann, X.

    1991-01-01

    Austenitic welds play an important role in reactor pressure vessels (RPV) both of pressurized water reactors (PWRs) and boiling water reactors (BWRs). These pressure retaining joints still pose problems for ultrasonic examination. During the last few years some progress has been achieved in overcoming these problems. The state of the art of the practical application is described in 'The Handbook on Ultrasonic Examination of Austenitic Welds' of the International Institute of Welding which was issued in 1986. This document was prepared by an international working group. Sulzer Brothers Limited, Winterthur Switzerland as manufacturer of both BWR and PWR RPVs has a long experience with ultrasonic examination of austenitic welds. In General Electric BWR RPVs the main recirculation line, the feedwater line and additional lines are connected to nozzles of the RPV with dissimilar metal welds joining stainless steel to carbon steel. The control rod drive (CRD) housings are attached to the RPV with complex weld joints. Framatome and Westinghouse RPVs also are provided with welds between the vessel and the main coolant line. This paper presents examples of ultrasonic approaches and examination results. Aspects of fabrication, preservice and inservice inspections are also covered. Complex weld joint design both from the point of material and geometry make ultrasonic examination sometimes extremely difficult but based on case by case investigations reliable solutions can often be realized. (orig.)

  15. Effects of moisture migration on shrinkage, pore pressure and other concrete properties

    International Nuclear Information System (INIS)

    Chapman, D.A.; England, G.L.

    1977-01-01

    This work investigates the uniaxial migration of moisture in long, upright, limestone concrete cylinders, sealed at the base and sides, and open at the top. The design represents a section through a concrete pressure vessel wall. The cylinders are subjected to a sustained temperature difference between their ends, with maximum temperatures between 105 0 C and 200 0 C. Readings of pore pressure, water content and temperature are taken at various positions along the axis of the cylinders. In one cylinder, transverse and longitudinal shrinkage readings are also recorded. The results for the cylinders show that moisture migration is away from the hot face of the specimens causing reduction in both pore pressure and water content values in this region. The moisture migration creates a drying front which moves slowly up the specimens. The rate at which this drying front, moves is influenced by the base temperature, the magnitude of temperature and pressure gradients and the coefficient of permeability of the concrete. Samples taken from the hot side of the drying front show a considerable increase in the coefficient of permeability, and Scanning Electron Microscope photographs of the microstructure show both a break-up and reduction in size of the hydration products. The experiments reported indicate that when the hot inner face temperature of a concrete pressure vessel is increased above 100 0 C, the drying rate inside the wall increases considerably, However, it is unlikely pressure vessels of the size currently in use will ever completely dry out. (Auth.)

  16. Guiding device for a manipulator mast for internal inspection of a reactor pressure vessel

    International Nuclear Information System (INIS)

    Seifert, W.; Schlueter, H.

    1977-01-01

    A remote-controlled supporting device centering a manipulator mast is described which is mounted and operated above a reactor pressure vessel under water in such a way that rotations and vertical movements necessary for the internal inspection of the pressure vessel remain possible. (RW) [de

  17. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    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 with...

  18. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    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 the...

  19. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and... METAL AND NONMETAL MINES Compressed Air and Boilers § 56.13015 Inspection of compressed-air receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels...

  20. Residual stresses in a weldment of pressure vessel steel

    International Nuclear Information System (INIS)

    Gott, K.E.

    1978-01-01

    A study was made of the distribution of residual stresses around a typical weld from a light water reactor pressure vessel by an X-ray double-exposure camera technique. So that the magnitude, sign, and distribution of the residual stresses were as similar as possible to those found in practice, a wide, full-thickness specimen of A533B Cl 1 steel containing a submerged-arc weld was stress-relief annealed. To obtain a three-dimensional distribution of the stresses the specimen was examined at different levels through the thickness. Following the removal of material by milling, the specimen surface was electropolished to free it from cold work. Corrections have been made to take into account specimen relaxation. To completely define the original stress system it is desirable also to measure the change in curvature on removing a layer of material. Unless this is done assumptions must be made which complicate the calculations unnecessarily. This became apparent after the experimental work was completed. In the centre of the plate the methods of correction which can be used are sensitive to errors in the measurements. The corrected results show that the dominant residual stress is perpendicular to the weld. It is positive at the surfaces and negative in the centre of the plate. The maximum value can reach the yield stress. The residual stresses in the weld metal can locally vary considerably: from 100 to 350N/mm 2 over a distance of 5mm. Such large variations have been found to coincide with the heat-affected zones of the individual weld runs. (author)

  1. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  2. Response and Damage Assessment of Reinforced Concrete Frames subject to Earthquakes

    DEFF Research Database (Denmark)

    Skjærbæk, Poul

    When civil engineering structures made of reinforced concrete (RC) such as some types of apartment buildings, hospitals, office buildings, bridges etc. are subjected to sufficiently high dynamic loads it is well known that some kind of damage will occur in the structure. The damage introduced in ...

  3. Local and Modal Damage Indicators for Reinforced Concrete Shear Frames Subject to Earthquakes

    DEFF Research Database (Denmark)

    Köylüoglu, H. U.; Nielsen, Søren R. K.; Abbott, J.

    Local, modal and overall damage indicators for reinforced concrete shear frames subject to seismic excitation are defined and studied. Each storey of the shear frame is represented by a Clough and Johnston hysteretic oscillator with degrading elastic fraction of the restoring force. The local max...

  4. Capacity assessment of concrete containment vessels subjected to aircraft impact

    International Nuclear Information System (INIS)

    Andonov, Anton; Kostov, Marin; Iliev, Alexander

    2015-01-01

    Highlights: • An approach to assess the containment capacity to aircraft impact via fragility curves is proposed. • Momentum over Area was defined as most suitable reference parameter to describe the aircraft load. • The effect of the impact induced damages on the containment pressure capacity has been studied. • The studied containment shows no reduction of the pressure capacity for the investigated scenarios. • The effectiveness of innovative protective structure against aircraft impact has been evaluated. - Abstract: The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept.

  5. Capacity assessment of concrete containment vessels subjected to aircraft impact

    Energy Technology Data Exchange (ETDEWEB)

    Andonov, Anton, E-mail: anton.andonov@mottmac.com; Kostov, Marin; Iliev, Alexander

    2015-12-15

    Highlights: • An approach to assess the containment capacity to aircraft impact via fragility curves is proposed. • Momentum over Area was defined as most suitable reference parameter to describe the aircraft load. • The effect of the impact induced damages on the containment pressure capacity has been studied. • The studied containment shows no reduction of the pressure capacity for the investigated scenarios. • The effectiveness of innovative protective structure against aircraft impact has been evaluated. - Abstract: The paper describes the procedure and the results from the assessment of the vulnerability of a generic pre-stressed containment structure subjected to a large commercial aircraft impact. Impacts of Boeing 737, Boeing 767 and Boeing 747 have been considered. The containment vulnerability is expressed by fragility curves based on the results of a number of nonlinear dynamic analyses. Three reference parameters have been considered as impact intensity measure in the fragility curve definition: peak impact force (PIF), peak impact pressure (PIP) and Momentum over Area (MoA). Conclusions on the most suitable reference parameter as well on the vulnerability of such containment vessels are drawn. The influence of the aircraft impact induced damages on the containment ultimate pressure capacity is also assessed and some preliminary conclusions on this are drawn. The paper also addresses a conceptual design of a protective structure able to decrease the containment vulnerability and provide a preliminary assessment of the applicability of such concept.

  6. Development of computational methods of design by analysis for pressure vessel components

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan; Wu Honglin

    2005-01-01

    Stress classification is not only one of key steps when pressure vessel component is designed by analysis, but also a difficulty which puzzles engineers and designers at all times. At present, for calculating and categorizing the stress field of pressure vessel components, there are several computation methods of design by analysis such as Stress Equivalent Linearization, Two-Step Approach, Primary Structure method, Elastic Compensation method, GLOSS R-Node method and so on, that are developed and applied. Moreover, ASME code also gives an inelastic method of design by analysis for limiting gross plastic deformation only. When pressure vessel components design by analysis, sometimes there are huge differences between the calculating results for using different calculating and analysis methods mentioned above. As consequence, this is the main reason that affects wide application of design by analysis approach. Recently, a new approach, presented in the new proposal of a European Standard, CEN's unfired pressure vessel standard EN 13445-3, tries to avoid problems of stress classification by analyzing pressure vessel structure's various failure mechanisms directly based on elastic-plastic theory. In this paper, some stress classification methods mentioned above, are described briefly. And the computational methods cited in the European pressure vessel standard, such as Deviatoric Map, and nonlinear analysis methods (plastic analysis and limit analysis), are depicted compendiously. Furthermore, the characteristics of computational methods of design by analysis are summarized for selecting the proper computational method when design pressure vessel component by analysis. (authors)

  7. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

  8. Stress categorization in nozzle to pressure vessel connections finite elements models; Categorizacao de tensoes em modelos de elementos finitos de conexoes bocal-vaso de pressao

    Energy Technology Data Exchange (ETDEWEB)

    Albuquerque, Levi Barcelos de

    1999-07-01

    The ASME Boiler and Pressure Vessel Code, Section III , is the most important code for nuclear pressure vessels design. Its design criteria were developed to preclude the various pressure vessel failure modes throughout the so-called 'Design by Analysis', some of them by imposing stress limits. Thus, failure modes such as plastic collapse, excessive plastic deformation and incremental plastic deformation under cyclic loading (ratchetting) may be avoided by limiting the so-called primary and secondary stresses. At the time 'Design by Analysis' was developed (early 60's) the main tool for pressure vessel design was the shell discontinuity analysis, in which the results were given in membrane and bending stress distributions along shell sections. From that time, the Finite Element Method (FEM) has had a growing use in pressure vessels design. In this case, the stress results are neither normally separated in membrane and bending stress nor classified in primary and secondary stresses. This process of stress separation and classification in Finite Element (FE) results is what is called stress categorization. In order to perform the stress categorization to check results from FE models against the ASME Code stress limits, mainly from 3D solid FE models, several research works have been conducted. This work is included in this effort. First, a description of the ASME Code design criteria is presented. After that, a brief description of how the FEM can be used in pressure vessel design is showed. Several studies found in the literature on stress categorization for pressure vessel FE models are reviewed and commented. Then, the analyses done in this work are presented in which some typical nozzle to pressure vessel connections subjected to internal pressure and concentrated loads were modeled with solid finite elements. The results from linear elastic and limit load analyses are compared to each other and also with the results obtained by formulae

  9. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

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

    Science.gov (United States)

    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.

  11. Renovation of the sealing planes of WWER-400 reactors pressure vessel

    International Nuclear Information System (INIS)

    Jablonicky, P.; Pilat, P.

    2007-01-01

    An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

  12. Fabrication techniques of metal liner used for pressure vessels made by composite material

    International Nuclear Information System (INIS)

    Takahashi, W.K.; Al-Qureshi, H.A.

    1982-01-01

    Different viable techniques for the manufacturing of metal liner used for pressure vessels are presented. The aim of these metal liner is to avoid the fluid leakage from the pressurized vessel and to serve as a mandreal to be wound by composite material. The studied techniques are described and the practical results are illustrated. Finally a comparative study of the manufacturing techniques is made in order to define the process that furnishes the metal liner with the best characteristics. The advantages offered by these type of pressure vessels when compared with the conventional metallic vessels, are also presented. (Author) [pt

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

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

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

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

    Science.gov (United States)

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

    2013-01-01

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

  15. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

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

    CERN Document Server

    Matthews, Clifford

    2010-01-01

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

  17. Creep Behavior of High-Strength Concrete Subjected to Elevated Temperatures.

    Science.gov (United States)

    Yoon, Minho; Kim, Gyuyong; Kim, Youngsun; Lee, Taegyu; Choe, Gyeongcheol; Hwang, Euichul; Nam, Jeongsoo

    2017-07-11

    Strain is generated in concrete subjected to elevated temperatures owing to the influence of factors such as thermal expansion and design load. Such strains resulting from elevated temperatures and load can significantly influence the stability of a structure during and after a fire. In addition, the lower the water-to-binder (W-B) ratio and the smaller the quantity of aggregates in high-strength concrete, the more likely it is for unstable strain to occur. Hence, in this study, the compressive strength, elastic modulus, and creep behavior were evaluated at target temperatures of 100, 200, 300, 500, and 800 °C for high-strength concretes with W-B ratios of 30%, 26%, and 23%. The loading conditions were set as non-loading and 0.33f cu . It was found that as the compressive strength of the concrete increased, the mechanical characteristics deteriorated and transient creep increased. Furthermore, when the point at which creep strain occurred at elevated temperatures after the occurrence of transient creep was considered, greater shrinkage strain occurred as the compressive strength of the concrete increased. At a heating temperature of 800 °C, the 80 and 100 MPa test specimens showed creep failure within a shrinkage strain range similar to the strain at the maximum load.

  18. Creep Behavior of High-Strength Concrete Subjected to Elevated Temperatures

    Directory of Open Access Journals (Sweden)

    Minho Yoon

    2017-07-01

    Full Text Available Strain is generated in concrete subjected to elevated temperatures owing to the influence of factors such as thermal expansion and design load. Such strains resulting from elevated temperatures and load can significantly influence the stability of a structure during and after a fire. In addition, the lower the water-to-binder (W–B ratio and the smaller the quantity of aggregates in high-strength concrete, the more likely it is for unstable strain to occur. Hence, in this study, the compressive strength, elastic modulus, and creep behavior were evaluated at target temperatures of 100, 200, 300, 500, and 800 °C for high-strength concretes with W–B ratios of 30%, 26%, and 23%. The loading conditions were set as non-loading and 0.33fcu. It was found that as the compressive strength of the concrete increased, the mechanical characteristics deteriorated and transient creep increased. Furthermore, when the point at which creep strain occurred at elevated temperatures after the occurrence of transient creep was considered, greater shrinkage strain occurred as the compressive strength of the concrete increased. At a heating temperature of 800 °C, the 80 and 100 MPa test specimens showed creep failure within a shrinkage strain range similar to the strain at the maximum load.

  19. Strength and deformation characteristics of reinforced concrete shell elements subjected to in-plane forces

    International Nuclear Information System (INIS)

    Aoyagi, Yukio; Yamada, Kazuie.

    1983-01-01

    Reactor containment vessels have been made of steel so far, but since it was decided to adopt a prestressed concrete vessel in the Tsuruga No. 2 plant of Japan Atomic Power Co., the construction of the containment vessels made of prestressed concrete and reinforced concrete has been studied by various electric power companies. However in Japan, there is no standard for the design and construction of concrete structures of this kind. In the standard of foreign countries used for reference, the basis of the stipulation concerning the aseismatic design of concrete containment vessels is not distinct. In this study, the clarification of the strength and deformation when RC vessels are subjected to seismic force only or to internal pressure and seismic force was aimed at, and the result of the loading test by one or two-direction in-plane forces on RC shell elements was examined. Based on this, the method of estimating the strength and deformation of RC shell elements was proposed. The orthogonal reinforcement was adopted, and the strength of shell elements was determined by the yielding of reinforcing bars. (Kako, I.)

  20. Simulation of reinforced concrete short shear wall subjected to cyclic loading

    International Nuclear Information System (INIS)

    Parulekar, Y.M.; Reddy, G.R.; Vaze, K.K.; Pegon, P.; Wenzel, H.

    2014-01-01

    Highlights: • Prediction of the capacity of squat shear wall using tests and analysis. • Modification of model of concrete in the softening part. • Pushover analysis using softened truss theory and FE analysis is performed. • Modified concrete model gives reasonable accurate peak load and displacement. • The ductility, ultimate load and also crack pattern can be accurately predicted. - Abstract: This paper addresses the strength and deformation capacity of stiff squat shear wall subjected to monotonic and pseudo-static cyclic loading using experiments and analysis. Reinforced concrete squat shear walls offer great potential for lateral load resistance and the failure mode of these shear walls is brittle shear mode. Shear strength of these shear walls depend strongly on softening of concrete struts in principal compression direction due to principal tension in other direction. In this work simulation of the behavior of a squat shear wall is accurately predicted by finite element modeling by incorporating the appropriate softening model in the program. Modification of model of concrete in the softening part is suggested and reduction factor given by Vecchio et al. (1994) is used in the model. The accuracy of modeling is confirmed by comparing the simulated response with experimental one. The crack pattern generated from the 3D model is compared with that obtained from experiments. The load deflection for monotonic loads is also obtained using softened truss theory and compared with experimental one

  1. The Combined Effects of Stress Concentration and Tensile Stresses from Autofrettage on the Life of Pressure Vessels

    Science.gov (United States)

    2017-02-01

    PRESSURE VESSELS E. Troiano G.N. Vigilante L.B. Smith J.H. Izzo February 2017 Approved for public...SUBTITLE THE COMBINED EFFECTS OF STRESS CONCENTRATION AND TENSILE STRESSES FROM AUTOFRETTAGE ON THE LIFE OF PRESSURE VESSELS 5a. CONTRACT NUMBER...Approved for public release; distribution is unlimited. 13. SUPPLEMENTARY NOTES 14. ABSTRACT Thick walled pressure vessels are often

  2. The measurement for level of marine high-temperature and high-pressure vessels

    International Nuclear Information System (INIS)

    Lin Jie.

    1986-01-01

    The various error factors in measurement for level of marine high-temperature and high-pressure vessels are anslysed. The measuring method of error self compensation and its simplification for land use are shown

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

    Science.gov (United States)

    Skow, Miles G.

    2014-01-01

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

  4. Research progress and recommendations on reactor pressure vessel integrity under hypothetical core melt down accident

    International Nuclear Information System (INIS)

    Yao Yangui; Ning Dong; Wu Zhiwei; Cao Ming; Xie Yongcheng; He Yinbiao; Yao Weida

    2013-01-01

    Background: It is very important to ensure the integrity of the reactor pressure vessel under core melt down accident. The high-temperature creep failure is the main failure mode of the reactor pressure vessel under core melt down accident. Purpose: This paper is to present an overview of research status and progress on high-temperature creep behavior of reactor pressure vessel considering the hypothetical core melt down scenario. Methods: Emphasis is placed on accomplished achievements in creep tests, scale model experiments and numerical simulation, and the domestic newly research productions on high-temperature creep behavior of reactor pressure vessel structure integrity. Conclusions: This paper also discusses the limitations of existing researches and indicates future research directions, such as multi-axis tensile tests, analysis of three-dimensional coupling temperature field, scaled model tests, and so on. (authors)

  5. Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1975-01-01

    Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

  6. Mechanical modelling of a structural performance of a pressure vessel submitted to the creep phenomenon

    International Nuclear Information System (INIS)

    Taroco, E.; Feijoo, R.A.; Monteiro, Edson; Freire, J.L.F.; Bevilacqua, L.; Miranda, P.E.V. de; Silveira, T.L. da

    1982-01-01

    A pressure vessel is analized using different mechanical models for the creep phenomenon. The numerical results obtained through these models enable us to recommend on the way verifications of creep damage accumulation is structures should be made. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  8. Remote controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Mueller, G.

    1990-01-01

    The first mechanised in-service inspection of the reactor pressure vessel on unit one of Eskom's Koeberg nuclear power station has been carried out. Since 1968 a whole range of manipulators to carry out remote controlled ultrasonic inspections of nuclear power station equipment has been developed. The inspection of a reactor pressure vessel using a central mast manipulator is described. 3 figs., 1 ill

  9. Possibility of use small size specimens for toughness evaluation of pressure vessel steels in surveillance programmes

    International Nuclear Information System (INIS)

    Petrequin, P.; Soulat, P.

    1979-01-01

    In order to caracterize fracture toughness evolution of pressure vessel steels under irradiation, two types of small size specimens able to be introduced in surveillance capsules were studied: small (12.5 mm thick) CT specimens and precracked Charpy. Size effects on determination of J1C on irradiated and unirradiated steel are studied using 12.5 and 25 mm thick CT specimens. Kid is measured using precracked Charpy specimens on several pressure vessels irradiated and unirradiated

  10. Evaluations of half-bead weld repair procedures with thick-wall pressure vessels

    International Nuclear Information System (INIS)

    Canonico, D.A.; Whitman, G.D.

    1978-01-01

    The results of research on the evaluation of the half-bead weld repair method for use on nuclear reactor components are reviewed from data obtained on thick-section test pieces and intermediate-size pressure vessels. Material properties, the magnitude of residual stresses and the structural behavior of flawed pressure vessels are being obtained to determine the adequacy of the weld repair method for application in thick-section components

  11. Overview of research trends and problems on Cr-Mo low alloy steels for pressure vessel

    International Nuclear Information System (INIS)

    Chi, Byung Ha; Kim, Jeong Tae

    2000-01-01

    Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding

  12. Monitoring the aging of pressure vessel steels by TEP measurements: Advantages and current limitations of the method

    International Nuclear Information System (INIS)

    Kleber, X.; Saillet, S.

    2011-01-01

    The TEP (Thermoelectric Power or Seebeck coefficient) characterizes the ability of a material to generate an electrical potential difference when the material is subjected to a heat flux. It can be defined from the Seebeck effect, which manifests itself in a circuit formed by two different metals subjected to a temperature gradient. The origin of the thermoelectric power is, as the resistivity, due to electronic phenomena occurring at the atomic scale in relation to the crystallographic structure of the material. TEP measurements are used to characterize small microstructural changes at the scale of crystal defects. The high sensitivity of TEP makes it an excellent probe able of detecting small changes in the microstructural state, including precipitation, dissolution of alloying elements, hardening and recovery after deformation. It has been shown recently that the TEP of pressure vessels steels was sensitive to irradiation, making this measurement technique a potential candidate for monitoring the aging of the pressure vessel steel. However, the first measurements on Charpy specimens of the EDF monitoring program (Pressure Vessel Surveillance Program) showed a strong negative effect of specimen geometry on the accuracy that can be achieved. In this paper we show what the origins of these inaccuracies are. From numerical simulation and finite element model, we describe the roles of the thermal contact resistance as well as the influence of the geometry of the blocks device. A model is proposed to overcome these negative effects. We also show the effect of the presence of heterogeneities in the material on the TEP measurement, and the importance of their localization. Finally, solutions are proposed to improve the device for measuring TEP on PVSP Charpy specimens. (authors)

  13. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

    2013-01-01

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  14. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Potirniche, Gabriel [Univ. of Idaho, Moscow, ID (United States); Barlow, Fred D. [Univ. of Idaho, Moscow, ID (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Rink, Karl [Univ. of Idaho, Moscow, ID (United States)

    2013-11-26

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  15. Influence of fuel loading on neutron field in WWER-440 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Stacho, M.; Slugen, V.; Farkas, G.; Sojak, S. [Department of Nuclear Physics and Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia)

    2010-07-01

    One of the limiting factors in terms of nuclear power plant lifetime is reactor pressure vessel neutron load. Neutron embrittlement as the most important ageing effect on the reactor pressure vessel is mainly caused by fast neutron spectra. The work is focused on mapping of neutron fields in the reactor pressure vessel of WWER-440/V-213 reactor using MCNP5 transport code. The calculation of neutron fields was performed using detailed full-core MCNP model of WWER-440 reactor developed at our department. Analysis of fuel loading pattern and burn-up influence on neutron flux density distribution in the reactor pressure vessel was realized. The fuel composition corresponds to fuel cycles of Bohunice and Mochovce nuclear power plants. The goal of this work was to improve the assessment of WWER-440 reactor pressure vessel radiation degradation and following evaluation possibility of its lifetime extension and comparison of neutron flux and neutron spectra in the most loaded place of reactor pressure vessel and surveillance specimen area. (authors)

  16. An Experimental Study of a Midbroken 2-Bay 6-Storey Reinforced Concrete Frame subject to Earthquakes

    DEFF Research Database (Denmark)

    Skjærbæk, P. S.; Taskin, B.; Kirkegaard, Poul Henning

    1997-01-01

    A 2-bay, 6-storey model test reinforced concrete frame (scale 1:5) subjected to sequential earthquakes of increasing magnitude is considered in this paper. The frame was designed with a weak storey, in which the columns are weakened by using thinner and weaker reinforcement bars. The aim of the w......A 2-bay, 6-storey model test reinforced concrete frame (scale 1:5) subjected to sequential earthquakes of increasing magnitude is considered in this paper. The frame was designed with a weak storey, in which the columns are weakened by using thinner and weaker reinforcement bars. The aim...... of the work is to study global response to a damaging strong motion earthquake event of such buildings. Special emphasis is put on examining to what extent damage in the weak storey can be identified from global response measurements during an earthquake where the structure survives, and what level...

  17. Studies on the behavior of Reinforced Concrete Short Column subjected to fire

    Directory of Open Access Journals (Sweden)

    Aneesha Balaji

    2016-03-01

    Full Text Available The paper investigates the axial capacity of reinforced columns exposed to fire. The simplified method namely 500 °C isotherm method explained in Eurocode 2 is used to assess the capacity of the column. Finite element software ANSYS is used to perform the thermal analysis. A set of numerical studies were carried out to quantify the effect of various parameters on short columns subjected to fire. The study is performed on columns of different cross-sections to investigate the effect of eight parameters, namely the thermal boundary conditions, grades of concrete, grades of steel, types of aggregate, distribution of reinforcement on column faces, concrete cover, load eccentricity and support conditions. The fire ratings based on various failure criteria are determined, and minimum rating is accepted as design fire rating. A simplified interaction curve to predict the failure of columns under fire subjected to axial load and uniaxial moment for square cross-sections was also developed.

  18. Effects of Rock Joints on Failure of Tunnels Subject to Blast Loading

    Science.gov (United States)

    2013-11-01

    13. SUPPLEMENTARY NOTES 14. ABSTRACT 15. SUBJECT TERMS 16. SECURITY CLASSIFICATION OF: 17. LIMITATION OF ABSTRACT 18. NUMBER OF...Modelling of joints and interfaces using the disturbed-state concept. International Journal for Numerical and Analytical Methods in Geomechanics , 16:623–653... Geomechan - ics, 11:345–362. Rashid, Y. (1968). Analysis of prestressed concrete pressure vessels. Nuclear Engineering and Design, 7:334–355. Schreyer

  19. Performance Evaluation of Waterproofing Membrane Systems Subject to the Concrete Joint Load Behavior of Below-Grade Concrete Structures

    Directory of Open Access Journals (Sweden)

    Jaeyoung Song

    2017-11-01

    Full Text Available Below-grade structures such as parking lots, underground subway tunnels, and basements are growing in scale and reaching deeper below-ground levels. In this type of environment, they become subject to higher water pressure. The concrete material of the structures is exposed to wet conditions for longer periods of time, which makes the proper adhesion of waterproofing membranes difficult. Joint movements from increased structural settlement, thermal expansion/shrinkage, and physical loads from external sources (e.g., vehicles make securing durable waterproofing challenging. While ASTM Guides, Korean Codes, and BS Practice Codes on below-grade waterproofing stress the importance of manufacturer specification for quality control, ensuring high quality waterproofing for the ever-changing scale of construction remains a challenge. This study proposes a new evaluation method and criteria which allow for the selection of waterproofing membranes based on specific performance attributes and workmanship. It subjects six different waterproofing membrane systems (installed on dry and wet surface conditioned mortar slab specimens with an artificial joint to different cyclic movement widths to 300 cycles in water to demonstrate that inadequate material properties and workmanship are key causes for leakages.

  20. Integrated surveillance specimen program for WWER-1000/V-320 reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Kytka, M.; Novosad, P.; Zdarek, J.

    2005-01-01

    Surveillance specimen programs play an important role in reactor pressure vessel lifetime assessment as they should monitor changes in pressure vessel materials, mainly their irradiation embrittlement. Standard surveillance programs in WWER-1000/V-320 reactor pressure vessels have some deficiencies resulting from their design-nonuniformity of neutron field and even within individual specimen sets, large gradient in neutron flux between specimens and containers, lack of neutron monitors in most of containers and no suitable temperature monitors. Moreover, location of surveillance specimens does not assure similar conditions as the beltline region of reactor pressure vessels. Thus, Modified surveillance program for WWER-1000/V-320C type reactors was designed and realized in two units of NPP Temelin, Czech Republic. In this program, large flat type containers are located on inner wall of reactor pressure vessel in the beltline region that assures their practically identical irradiation conditions with critical vessel materials. These containers with inner dimensions of 210 x 300 mm have two layers of specimens; using inserts (10 x 10 x 14 mm) instead of fully Charpy size specimens allows irradiation of materials from several pressure vessels at once in one container. This design advantage has been used for the creation of the Integrated Surveillance Program for several WWER-1000 units-Temelin 1 + 2, Belene (Bulgaria), Rovno 3 + 4, Khmelnick 2, Zaporozhie 6 (Ukraine) and Kalinin 3 (Russia). Irradiation of these archive materials together with the IAEA reference steel JRQ (of ASTM A 533-B type) and reference steel VVER-1000 will allow to compare irradiation embrittlement of these materials and to obtain more reliable and objective results as no reliable predictive formulae exist up to no due to a higher content of nickel in welds. Irradiation of specimens from cladding region will help in the evaluation of resistance of pressure vessels against PTS regimes. (authors)

  1. Residual characteristic properties of ternary blended steel fibre reinforced concrete subjected to sustained elevated temperature

    Directory of Open Access Journals (Sweden)

    Sinha Deepa A.

    2013-09-01

    Full Text Available To study the behavior of ternary blended steel fibre reinforced concrete when subjected to 800 Deg.C and 1000 Deg.C for 3 hours. It has been found that the ternary blended steel fibre reinforced concrete containing (FA+GGBFS and (FA+MK offer higher resistance to sustained elevated temperatures upto 800 Deg.C, where as the blend containing (FA+SF does not offer any resistance at this temperature. The study reveals that the blend containing (FA+GGBFS and (FA+MK gives highest resistance at replacement levels of (10+20 and (15+15 respectively at sustained exposure to 800 Deg.C.

  2. Evaluation of HFIR (High Flux Isotope Reactor) pressure-vessel integrity considering radiation embrittlement

    Energy Technology Data Exchange (ETDEWEB)

    Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K. (eds.)

    1988-04-01

    The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of approx.10/sup 4/ less), that is, a rate effect.

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

    International Nuclear Information System (INIS)

    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

  4. Creep failure of a reactor pressure vessel lower head under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pilch, M.M.; Ludwigsen, J.S.; Chu, T.Y. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R. [Anatech, San Diego, CA (United States)

    1998-08-01

    A severe accident in a nuclear power plant could result in the relocation of large quantities of molten core material onto the lower head of he reactor pressure vessel (RPV). In the absence of inherent cooling mechanisms, failure of the RPV ultimately becomes possible under the combined effects of system pressure and the thermal heat-up of the lower head. Sandia National Laboratories has performed seven experiments at 1:5th scale simulating creep failure of a RPV lower head. This paper describes a modeling program that complements the experimental program. Analyses have been performed using the general-purpose finite-element code ABAQUS-5.6. In order to make ABAQUS solve the specific problem at hand, a material constitutive model that utilizes temperature dependent properties has been developed and attached to ABAQUS-executable through its UMAT utility. Analyses of the LHF-1 experiment predict instability-type failure. Predicted strains are delayed relative to the observed strain histories. Parametric variations on either the yield stress, creep rate, or both (within the range of material property data) can bring predictions into agreement with experiment. The analysis indicates that it is necessary to conduct material property tests on the actual material used in the experimental program. The constitutive model employed in the present analyses is the subject of a separate publication.

  5. Minutes of the Twelfth LWR pressure vessel surveillance dosimtery improvement program meeting

    International Nuclear Information System (INIS)

    1989-01-01

    The 1983 Twelfth Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) Meeting, which was held October 24-28, 1983. Sections 1 through 14 of this report provide documentation of agreements, commitments, and reports that are subject to the approval and concurrence of the participating laboratories and supporting agencies and organizations. Attachment No. 1 provides information on the preparation of a number of NUREG publications that will document the results of various aspects of the LWR-PV-SDIP. For each NUREG publication, a tentative ''Table of Contents'' is provided in addition to suggested interlaboratory writing assignments and camera-ready copy contribution due dates, as appropriate. Attachment No. 2 provides information on planning for the Fifth ASTM-EURATOM Symposium. Attachment No. 3 provides information on an ASTM press release about an MPC-6 meeting and dpa and E > 1 MeV exposure parameters. Attachments No. 4 and 5 provide copies of two LWR-PV-SDIP related papers presented at the Eleventh WRSR Information Meeting, October 24-28, 1983

  6. Large inelastic deformation analysis of steel pressure vessels at high temperature

    International Nuclear Information System (INIS)

    Ikonen, K.

    2001-01-01

    This publication describes the calculation methodology developed for a large inelastic deformation analysis of pressure vessels at high temperature. Continuum mechanical formulation related to a large deformation analysis is presented. Application of the constitutive equations is simplified when the evolution of stress and deformation state of an infinitesimal material element is considered in the directions of principal strains determined by the deformation during a finite time increment. A quantitative modelling of time dependent inelastic deformation is applied for reactor pressure vessel steels. Experimental data of uniaxial tensile, relaxation and creep tests performed at different laboratories for reactor pressure vessel steels are investigated and processed. An inelastic deformation rate model of strain hardening type is adopted. The model simulates well the axial tensile, relaxation and creep tests from room temperature to high temperature with only a few fitting parameters. The measurement data refined for the inelastic deformation rate model show useful information about inelastic deformation phenomena of reactor pressure vessel steels over a wide temperature range. The methodology and calculation process are validated by comparing the calculated results with measurements from experiments on small scale pressure vessels. A reasonably good agreement, when taking several uncertainties into account, is obtained between the measured and calculated results concerning deformation rate and failure location. (orig.)

  7. Evaluation of detectable defect size for inner defect of pressure vessel using laser speckle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeoung Suk; Seon, Sang Woo; Choi, Tae Ho; Kang, Chan Geun; Na, Man Gyun; Jung, Hyun Chul [Chsoun University, Gwangju (Korea, Republic of)

    2014-04-15

    Pressure vessels are used in various industrial fields. If a defect occurs on the inner or outer surface of a pressure vessel, it may cause a massive accident. A defect on the outer surface can be detected by visual inspection. However, a defect on the inner surface is generally impossible to detect with visual inspection. Nondestructive testing can be used to detect this type of defect. Laser speckle shearing interferometry is one nondestructive testing method that can optically detect a defect; its advantages include noncontact, full field, and real time inspection. This study evaluated the detectable size for an internal defect of a pressure vessel. The material of the pressure vessel was ASTM A53 Gr.B. The internal defect was detected when the pressure vessel was loaded by internal pressure controlled by a pneumatic system. The internal pressure was controlled from 0.2 MPa to 0.6 MPa in increments of 0.2 MPa. The results confirmed that an internal defect with a 25 % defect depth could be detected even at 0.2 MPa pressure variation.

  8. Measuring method and device for thickness of pad welded joint of reactor pressure vessel

    International Nuclear Information System (INIS)

    Ara, Katsuyuki; Nakajima, Nobuya; Ebine, Noriya.

    1995-01-01

    A magnetic yoke having an optional magnetic path length and a magnetic path cross section is disposed in close contact with or adjacent to the surface of the pad welded joint of a reactor pressure vessel, to form a magnetic path by the magnetic yoke and the reactor pressure vessel. Then, the magnetic path yoke is magnetized to measure a distribution of a magnetic field generated on the surface of the pad welded joint or its vicinity to which the magnetic yoke is adapted to be in close contact or come close. Since the geometrical dimensions and the magnetic performances of a material of the magnetic yoke and the pressure vessel are previously determined and, the measured magnetic distribution changes only by the thickness of the pad welded joint, the thickness of the pad welded joint of the pressure vessel can be determined. Accordingly, the measuring method of the present invention can measure the thickness of the pad welded joint of the pressure vessel at a desired practical accuracy and at a resolution power of, for example, 0.1mm relative to the thickness of the seat welded joint of 5 to 10mm. (N.H.)

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

    International Nuclear Information System (INIS)

    Kitagawa, Hideo; Hisada, Toshiaki

    1979-01-01

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

  10. Thermal analysis of GFRP-reinforced continuous concrete decks subjected to top fire

    Science.gov (United States)

    Hawileh, Rami A.; Rasheed, Hayder A.

    2017-12-01

    This paper presents a numerical study that investigates the behavior of continuous concrete decks doubly reinforced with top and bottom glass fiber reinforced polymer (GFRP) bars subjected to top surface fire. A finite element (FE) model is developed and a detailed transient thermal analysis is performed on a continuous concrete bridge deck under the effect of various fire curves. A parametric study is performed to examine the top cover thickness and the critical fire exposure curve needed to fully degrade the top GFRP bars while achieving certain fire ratings for the deck considered. Accordingly, design tables are prepared for each fire curve to guide the engineer to properly size the top concrete cover and maintain the temperature in the GFRP bars below critical design values in order to control the full top GFRP degradation. It is notable to indicate that degradation of top GFRP bars do not pose a collapse hazard but rather a serviceability concern since cracks in the negative moment region widen resulting in simply supported spans.

  11. 2D Idealisation of Hollow Reinforced Concrete Beams Subjected to Combined Torsion, Bending and Shear

    Directory of Open Access Journals (Sweden)

    A.S. Al-Nuaimi

    2005-12-01

    Full Text Available This paper presents a finite element model for idealisation of reinforced concrete hollow beams using 2D plane elements. The method of ensuring compatibility between the plates using two-dimensional model to analyze this type of structures is discussed. Cross-sectional distortion was minimised by incorporating end diaphragms in the FE model. Experimental results from eight reinforced concrete hollow beams are compared with the non-linear predictions produced by a 2D in-house FE program. The beam dimensions were 300x300 mm cross section with 200x200 mm hollow core and 3800 mm long. The beam ends were filled with concrete to form solid end diaphragms to prevent local distortion. The beams were subjected to combined bending, torsion and shear. It was found that the two-dimensional idealisation of hollow beams is adequate provided that compatibility of displacements between adjoining plates along the line of intersection is maintained and the cross-sectional distortion is reduced to minimum. The results from the 2D in-house finite element program showed a good agreement with experimental results.

  12. Dynamic Behavior of a Prestressed Concrete Bridge with a Switching Crack Subjected to Moving Trains

    Directory of Open Access Journals (Sweden)

    Chunyu Fu

    2016-01-01

    Full Text Available Concrete bridges chronically exposed to the natural environment are vulnerable to cracking. Under prestressing forces, the crack may close, but it will open under large dynamic loads, such as heavy trains. So the crack state can switch during the vibration. This paper investigates the nonlinear dynamic behavior of the prestressed concrete bridge with such a switching crack subjected to moving trains. Firstly, the finite element method is adopted to formulate the motion equations of train-bridge system while the crack state remains constant. Then the displacement components are analyzed to investigate the effect of static loads at the switching instant. Finally, an iterative algorithm is adopted to calculate the nonlinear responses. The proposed method is verified by the calculation of an actual prestressed concrete bridge, and the results show that the crack switching can increase the peak of the bridge displacement during the vibration, cause an obvious nonlinearity between the displacement and the vehicle mass, and make the displacement become more variable with the train velocity.

  13. Size effect studies of the creep behaviour of a pressure vessel steel at temperatures from 700 to 900 deg. C

    International Nuclear Information System (INIS)

    Krompholz, Klaus; Kalkhof, Dietmar

    2002-01-01

    The study of size and scale effects in plastic flow and failure is of great importance for the evaluation of the pressure vessel behaviour in severe accidents. The forged reactor pressure vessel material 20MnMoNi55, material number DIN 1.6310 (heat number 69906) was subjected to creep investigations under constant tension load using geometrically similar smooth specimens covering a scaling factor of 4. The tests were performed at 700, 800 and 900 deg. C in an inert atmosphere. The temperatures cover the phase transformation regime. The mechanical stress varied from 10 to 30 MPa, depending on the temperatures. The creep curves and characteristic data are size dependent to a varying degree, depending on the stress and temperature level. The size effect is largest at 700 deg. C at the lowest stress level and it implies that the above mentioned times are considerably longer for the small specimens. With increasing stress the size of the effect decreases and at the highest stress a reversal of the effect may occur. The type of fracture is strongly temperature dependent

  14. An integrated CAD/CAM system for CNG pressure vessel manufactured by deep drawing and ironing operation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joon Hong; Kim, Chul; Choi, Jae Chan [Pusan National Univ., Pusan (Korea, Republic of)

    2004-06-01

    The fiber reinforced composite material is widely used in the multi-industrial field because of their high specific modulus and specific strength. It has two main merits which are to cut down energy by reducing weight and to prevent explosive damage proceeding to the sudden bursting which is generated by the pressure leakage condition. Therefore, pressure vessels using this composite material can be applied in the field such as defence industry and aerospace industry. In this paper, for nonlinear finite element analysis of E-glass/epoxy filament winding of composite vessel subjected to internal pressure, the standard interpretation model is developed by using the ANSYS with AutoLISP and ANSYS APDL languages, general commercial software, which is verified as useful characteristic of the solution. Among the modules of the system, both the process planning module for carrying out the process planning of filament wound composite pressure vessel and the autofrettage process module for obtaining higher residual stress will minimize trial and error and reduce the period for developing new products. The system can serve as a valuable system for experts and as a dependable training aid for beginners.

  15. FINITE ELEMENT ANALYSIS OF THE BEHAVIOUR OF REINFORCED CONCRETE COLUMNS CONFINED BY OVERLAPPING HOOPS SUBJECTED TO RAPID CONCENTRIC LOADING

    Directory of Open Access Journals (Sweden)

    Xiang Zeng

    2017-12-01

    Full Text Available The strain rate sensitivity of concrete material was discovered approximately one hundred years ago, and it has a marked effect on the behaviour of concrete members subjected to dynamic loadings such as strong earthquake and impact loading. Because of the great importance of the confined reinforced concrete (RC columns in RC structures, the dynamic behaviour of the columns induced by the strain rate effect has been studied, but only few experiments and analyses have been conducted. To investigate the behaviour of overlapping hoop-confined square reinforced normal-strength concrete columns, considering the strain rate effect at a strain rate of 10-5/sec to 10-1/sec induced by earthquake excitation, an explicit dynamic finite element analysis (FEA model was developed in ABAQUS to predict the behaviour of confined RC columns subjected to the rapid concentric loading. A locally modified stress-strain relation of confined concrete with the strain rate sensitivity of the concrete material and the confining effect of overlapping hoops were proposed to complete the simulation of the dynamic behaviour of concrete with the concrete plastic-constitutive model in ABAQUS. The finite element predictions are consistent with the existing test results. Based on the FEA model, a parametric investigation was conducted to capture more information about the behaviour of confined RC columns under varying loading rates.

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

    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

  18. Investigation of the safety testing of extraordinary thick steel material for pressure vessels

    International Nuclear Information System (INIS)

    Vajmelka, K.

    1978-10-01

    With increasing size of nuclear reactors it was necessary to increase the wall thickness of the reactor pressure vessels. So, for example the wall thickness of the pressure vessel of the LWR pressurized water reactor type with 1.200.000 kW performance is 250 mm. The fabrication of these extraordinary thick steel plates is accurately carried out, nevertheless it is very important to control the characteristics of strength. For this reason extraordinarily thick steel plates of forged manganese-molybdenum-nickel steel produced in Japan and used for the production of reactor pressure vessels was utilized. The aim of this investigation is to know if and how the K sub(IR)- values correspond to the actual regulation for the prevention of brittle fracture and to determine the characteristics. (orig./RW) [de

  19. Aging results for PRD 49 III/epoxy and Kevlar 49/epoxy composite pressure vessels

    Science.gov (United States)

    Hamstad, M. A.

    1983-01-01

    Kevlar 49/epoxy composite is growing in use as a structural material because of its high strength-to-weight ratio. Currently, it is used for the Trident rocket motor case and for various pressure vessels on the Space Shuttle. In 1979, the initial results for aging of filament-wound cylindrical pressure vessels which were manufactured with preproduction Kevlar 49 (Hamstad, 1979) were published. This preproduction fiber was called PRD 49 III. This report updates the continuing study to 10-year data and also presents 7.5-year data for spherical pressure vessels wound with production Kevlar 49. For completeness, this report will again describe the specimens of the original study with PRD 49 as well as specimens for the new study with Kevlar 49.

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

    Science.gov (United States)

    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.

  1. New paradigm for prediction of radiation life-time of reactor pressure vessel

    International Nuclear Information System (INIS)

    Kotrechko, S.A.; Meshkov, Yu.Ya.; Neklyudov, I.M.; Revka, V.N.

    2011-01-01

    New paradigm for prediction of radiation life-time of reactor pressure vessel is presented. Equation for limiting state of reactor pressure vessel wall with crack-like defect is obtained. It is exhibited that the value of critical fluence Φ c may be determined not by shift of critical temperature of fracture of surveillance specimen, which is indirect characteristic, but by direct method, namely, by the condition of initiation of brittle fracture of irradiated metal ahead of a crack in RPV wall. Within the framework of engineering version of LA to fracture the technique for Φ c ascertainment is developed. Prediction of Φ c for WWER pressure vessels demonstrates potentialities of this technique.

  2. Investigation of the failure of a reactor pressure vessel by plastic instability

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1994-01-01

    A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    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

  4. 46 CFR 54.01-2 - Adoption of division 1 of section VIII of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

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

  5. Strength Prediction and Failure Modes of Concrete Specimens Subjected to the Split Test

    DEFF Research Database (Denmark)

    Hoang, Linh Cao; Andersen, M.E.; Hansen, N.T.

    2014-01-01

    This paper deals with modelling and test of concrete specimens subjected to the Brazilian split test. Based on the fictitious crack concept, a simple model for the crack propagation process in the splitting plane is developed. From the model, it is possible to determine the distribution of residual...... tensile strength as crack propagation take place. The residual tensile strength is thereafter used in a rigid plastic analysis of the splitting failure. Based on this combined approach, the ultimate load may either be governed by crack propagation or by a plastic failure, which then terminates the crack...

  6. Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

    A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

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

    Science.gov (United States)

    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.

  8. Programmable - logic equipment for ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Haniger, L.

    1980-01-01

    Two alternatives are presented of programmable logic corresponding to the 2nd generation of the apparatus for performing periodic ultrasonic inspections of power reactor pressure vessels and a solution is outlined of inspecting the circumferential weld on the pressure vessel head. The apparatus will allow using any measuring head taken into consideration for operational inspection. Command words are taken from a punched type reader. Czechoslovak made RAM memories are used. The algorithm of instrument function is supposed to be controlled by a microprocessor as soon as necessary preconditions for this technology are created in Czechoslovakia

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Trenberth, R.

    1987-12-01

    Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

  11. Advanced nickel/hydrogen dependent pressure vessel (DPV) cell and battery concepts

    Science.gov (United States)

    Caldwell, Dwight B.; Fox, C. L.; Miller, L. E.

    The dependent pressure vessel (DPV) nickel/hydrogen (NiH 2) design is being developed by Eagle-Picher Industries, Inc. (EPI) as an advanced battery for military and commercial aerospace and terrestrial applications. The DPV cell design offers high specific energy and energy density as well as reduced cost, while retaining the established individual pressure vessel (IPV) technology, flight heritage and database. This advanced DPV design also offers a more efficient mechanical, electrical and thermal cell and battery configuration and a reduced parts count. The DPV battery design promotes compact, minimum volume packaging and weight efficiency, and delivers cost and weight savings with minimal design risks.

  12. Mechanical properties of two-way different configurations of prestressed concrete members subjected to axial loading

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chaobi; Chen, Jian Yun; Xu, Qiang; Li, Jing [School of Civil and Hydraulic Engineering, Dalian University of Technology, Dalian (China)

    2015-08-15

    In order to analyze the mechanical properties of two-way different configurations of prestressed concrete members subjected to axial loading, a finite element model based on the nuclear power plant containments is demonstrated. This model takes into account the influences of different principal stress directions, the uniaxial or biaxial loading, and biaxial loading ratio. The displacement-controlled load is applied to obtain the stress-strain response. The simulated results indicate that the differences of principal stress axes have great effects on the stress-strain response under uniaxial loading. When the specimens are subjected to biaxial loading, the change trend of stress with the increase of loading ratio is obviously different along different layout directions. In addition, correlation experiments and finite element analyses were conducted to verify the validity and reliability of the analysis in this study.

  13. Prediction model for carbonation depth of concrete subjected to freezing-thawing cycles

    Science.gov (United States)

    Xiao, Qian Hui; Li, Qiang; Guan, Xiao; Xian Zou, Ying

    2018-03-01

    Through the indoor simulation test of the concrete durability under the coupling effect of freezing-thawing and carbonation, the variation regularity of concrete neutralization depth under freezing-thawing and carbonation was obtained. Based on concrete carbonation mechanism, the relationship between the air diffusion coefficient and porosity in concrete was analyzed and the calculation method of porosity in Portland cement concrete and fly ash cement concrete was investigated, considering the influence of the freezing-thawing damage on the concrete diffusion coefficient. Finally, a prediction model of carbonation depth of concrete under freezing-thawing circumstance was established. The results obtained using this prediction model agreed well with the experimental test results, and provided a theoretical reference and basis for the concrete durability analysis under multi-factor environments.

  14. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    Energy Technology Data Exchange (ETDEWEB)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

  15. Stress corrosion cracking of ferritic reactor pressure vessel steels under boiling water reactor conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.

    2001-01-01

    The stress corrosion cracking (SCC) behaviour of low-alloy reactor pressure vessel (RPV) steels in oxygenated high-temperature water and its relevance to boiling water reactor (BWR) power operation, in particular its possible effect on both, RPV structural integrity and safety, has been a subject of controversial discussions for many years. The SCC crack growth behaviour of different RPV steels under simulated BWR/NWC conditions was therefore characterized by constant load and ripple load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. Modern high-temperature water loops, online crack growth monitoring (DCPD) and fractographical analysis by scanning electron microscopy were used to quantify the cracking response. It is concluded that there is no susceptibility to sustained SCC crack growth at temperatures around 288 C under purely static loading, as long as small-scale-yielding conditions prevail at the crack tip and the water chemistry is maintained within current BWR/NWC operational practice (EPRI water chemistry guidelines). However, sustained, fast SCC (with respect to operational time scales) cannot be excluded for faulted water chemistry conditions (EPRI Action Level 3) and/or for highly stressed specimens, either loaded near to K IJ or with a high degree of plasticity in the remaining ligament. The conservative character of the 'BWR VIP 60 Disposition Lines 1 and 2' for SCC crack growth in low-alloy steels has been confirmed by this study for 288 C and RPV base material. Preliminary results indicate, that these disposition lines may be significantly or slightly exceeded (even in steels with low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 - 250 C) in RPV materials, which show a distinct susceptibility to Dynamic Strain Ageing (DSA). (orig.)

  16. Experimental study of thin square slabs, embedded on all four sides (pressure vessel floors)

    International Nuclear Information System (INIS)

    Valentin, A.

    1967-03-01

    This work was designed to test experimentally the mechanical strength of large thin plates made of non-magnetic stainless steel and subjected to transverse forces; these plates are intended to be used as the flooring of pressure vessels. The characteristics are as follows: Test 1: type: NS 22 S - Ugine; overall cross-section mm x mm: 1497 x 1555,5; thickness mm: 3 Test 2: type: Fluginox 130 - Ugine; overall cross-section mm x mm: 2100 x 2100; thickness mm: 3 Test 3: type: 832 MVR - Avesta; overall cross-section mm x mm: 2090 x 2090; thickness mm: 5. The plates were mounted on frames, each pillar of which had been previously pre-stressed, at its centre point, with a force of 300 kg (test 1), 600 kg (test 2), 800 kg (test 3). The experiments showed that the amounts of flexion at the centre w were related to the loads q by equations of the type: qa 4 /Dh = w/h[A+B(w/h) 2 ] where 2a, h and D represent respectively the length of a side, the thickness and the rigidity under flexion of the plates held on the four sides. The coefficients A and B have been determined for different cases; the equations obtained are discussed. It is shown in particular that prestressing of the pillars is beneficial but that the movement of the plates at their edges cannot be considered to be non-existent with the set-ups used. (author) 1967 [fr

  17. Ultrasonic stress evaluation through thickness of a stainless steel pressure vessel

    International Nuclear Information System (INIS)

    Javadi, Yashar; Pirzaman, Hamed Salimi; Raeisi, Mohammadreza Hadizadeh; Najafabadi, Mehdi Ahmadi

    2014-01-01

    This paper investigates ultrasonic method in stress measurement through thickness of a pressure vessel. Longitudinal critically refracted (L CR ) waves are employed to measure the welding residual stresses in a vessel constructed from austenitic stainless steel 304L. The acoustoelastic constant is measured through a hydro test to keep the pressure vessel intact. Hoop and axial residual stresses are evaluated by using different frequency range of ultrasonic transducers. The welding processes of vessel shell and caps are simulated by a 3D finite element (FE) model which is validated by hole-drilling method. The residual stresses calculated by FE simulation are then compared with those obtained from the ultrasonic measurement while a good agreement is observed. It is demonstrated that the residual stresses through thickness of the stainless steel pressure vessel can be evaluated by combining FE and L CR method (known as FEL CR method). - Highlights: • The main goal is ultrasonic evaluation of through thickness stresses. • Welding processes of a stainless steel pressure vessel are modelled by FE. • The hole-drilling method is used to validate the FE results. • Residual stresses are measured by four different series of ultrasonic transducers. • The comparison between ultrasonic and FE results show an acceptable agreement

  18. Modelling creep of pressure vessels with thermal gradients using Theta projection data

    International Nuclear Information System (INIS)

    Law, M.; Payten, W.; Snowden, K.

    2002-01-01

    Pressure vessels are often exposed to through-wall temperature gradients. Thermal stresses occur in addition to pressure stresses. The resulting creep response is calculated using the Theta projection creep algorithm within a finite element code. It was found that the stress and temperature dependence of the creep response may lead to complex stress evolution

  19. Temperature and stress distribution in pressure vessel by the boundary element method

    International Nuclear Information System (INIS)

    Alujevic, A.; Apostolovic, D.

    1990-01-01

    The aim of this paper is to demonstrate the applicability of boundary element method for the solution of temperatures and thermal stresses in the body of reactor pressure vessel of the NPP Krsko . In addition to the theory of boundary elements for thermo-elastic continua (2D, 3D) results are given of a numerically evaluated meridional cross-section. (author)

  20. Stress analysis in pressure vessels by mixed finite element methods taking into account shear deformation

    International Nuclear Information System (INIS)

    Franca, L.P.; Toledo, E.M.; Loula, A.F.D.; Garcia, E.L.M.

    1988-12-01

    A new finite element method is employed to approximate axisymmetric shell problems. This formulation enhances stability and accuracy, from thin to moderately thick shells, compared to the correspondent Galerkin finite element approximations. Numerical results illustrate the good performance of the present method on some typical pressure vessels aplications. (author) [pt

  1. Vertical CO2 release experiments from a 1 liter high pressure vessel

    NARCIS (Netherlands)

    Hulsbosch-Dam, C.; Jong, A. de; Zevenbergen, J.; Peeters, R.; Spruijt, M.

    2013-01-01

    High-speed CO2 release experiments from a pressurized vessel have been conducted for a range of storage pressures nozzle diameters. The experiments are conducted to validate and develop models for accidental releases of CO2 during transport. In cooperation with DNV-KEMA lab (Groningen, The

  2. Chemical methods for the use of niobium from pressure vessel cladding as a fast neutron dosimeter

    International Nuclear Information System (INIS)

    Karnani, Hari

    1986-08-01

    the steel samples from the cladding of a pressure vessel of an operating nuclear power reactor were obtained by scraping. The cladding material of the pressure vessel contained about 0.5 % niobium. It was desired to use the niobium as a dosimeter for estimating fast fluences at the pressure vessel. The weak radiation from the reaction product 93m Nb cannot be measured in the presence of other elements and interfering activities. A method was developed to separate niobium from other metals present; the concentration and yield of niobium were determined spectrophotometrically. The irradiated niobium was electrodeposited from aqueous solutions on copper discs. The amount of the deposited niobium was determined by a radiochemical method which makes use of its own radioactivity - measured with a liquid scintillation counter - and the known starting mass of niobium. It was possible to determine the deposited niobium masses (5 to 200 microgram) with a desired degree of accuracy. The absolute emission rate of X-rays could then be measured without any self-absorption or interference from other activities. The mass of niobium on each preparate and its X-ray emission rate, later on, were used as basic experimental data for the estimation of last neutron doses at the pressure vessel

  3. PWR pressure vessel life management French approach for integrity assessment and maintenance strategy

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), 93 - Saint-Denis (France); Moinereau, D. [Electricite de France, 77 - Moret sur Loing (France). Direction des Etudes et Recherches; Pichon, C.; Faidy, C.; Ternon-Morin, F. [Electricite de France (EDF), 69 - Villeurbanne (France). SEPTEN; Brillaud, C. [Electricite de France (EDF), 37 - Avoine (France)

    1998-07-01

    This conference deals with the studies carried out in France to justify of a PWR pressure vessel lifetime of at least 40 years. The results of the irradiation surveillance programs and of the fluences evaluation in the end of life are given as well as the EDF maintenance strategy. (O.M.)

  4. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and...-UNDERGROUND METAL AND NONMETAL MINES Compressed Air and Boilers § 57.13015 Inspection of compressed-air receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure...

  5. Prestressed and reinforced concrete containments. Analysis - design - construction

    International Nuclear Information System (INIS)

    Schnellenbach, G.

    1975-01-01

    Nuclear reactors performing in the German Federal Republic to date were supplied with steel containments. The first reinforced concrete and prestressed concrete containments, respectively, are going to be used for the nuclear power plants Kalkar and Gundremmingen (KRB II) as well as for the HTR plant. Because of their function and nature of loading these structures, similarly to the prestressed concrete reactor pressure vessels, belong to the special structures of civil engineering. Yet, they are substantially different from the prestressed concrete reactor pressure vessels. The problems connected with analysis, design, and construction of these structures are new as well. (orig.) [de

  6. Application of shearography to crack detection in concrete structures subjected to traffic loading

    Science.gov (United States)

    Muzet, V.; Blain, P.; Przybyla, D.

    2010-09-01

    Early detection of defects in concrete structures, such as bridges or dams, is essential to optimize the maintenance of civil engineering facilities. Optical methods constitute non-destructive means of control and measurement but they are generally confined in laboratories where both the setup and environnement are controlled. The method of shearography is especially well adapted to detect damages due to both its capacity to distinctly visualize strain concentration zones and its robustness. The experimental set-up is relatively compact, which enables to examine an extensive surface area by simply moving the shearographic head. In this paper, the application of this methodology for the detection of cracks is presented on concrete samples and circulated outside concrete structures. Due to its sensitivity to strain concentration, shearography is able to detect structural cracks, even when they were not through-cracks. Operational implementation is made on two circulated structures with experts in manual cracks detection. No stimulation device is used. In the first structure, cracks are detected on the bridge deck and on the bridge abutment. In the second structure, cracks on the intrados of the bridge deck are detected and also beginning of cracks which have not been detected by the visual inspection. Different areas are scanned and the results are in agreement with the visual inspection. This technique enables detecting cracks on structures subjected to traffic load. The natural loading of an engineering structure, i.e. the rolling traffic it bears, proves well suited for cracks detection by means of shearography, provided traffic patterns are regular enough.

  7. Joining dissimilar stainless steels for pressure vessel components

    International Nuclear Information System (INIS)

    Zheng Sun; Huai-Yue Han

    1994-01-01

    A series of studies was carried out to examine the weldability and properties of dissimilar steel joints between martensitic and austenitic stainless steels - F6NM (OCrl3Ni4Mo) and AISI 347, respectively. The weldability tests included weld thermal simulation of the martensitic steel for investigating the influence of weld thermal cycles and post-weld heat treatment (PWHT) on the mechanical properties of the heat-affected zone (HAZ); implant testing for examining the tendency for cold cracking of martensitic steel; rigid restraint testing for determining hot crack susceptibility of the multi-pass dissimilar steel joints. The joints were subjected to various mechanical tests including a tensile test, bending test and impact test at various temperatures, as well as slow strain-rate test for examining the stress corrosion cracking tendency in the simulated environment of a primary circuit of a PWR. Based on the weldability tests, a welding procedure - tungsten inert gas (TIG) welding for root passes with HNiCrMo-2B wire followed by manual metal arc (MMA) welding using coated electrode ENiCrFe-3B - was developed and a PWHT at 600 deg C/2h was recommended. Furthermore, the welding of tube/tube joints between these dissimilar steels is described. (21 refs., 11 figs., 14 tabs.)

  8. Optimization of Composite Material System and Lay-up to Achieve Minimum Weight Pressure Vessel

    Science.gov (United States)

    Mian, Haris Hameed; Wang, Gang; Dar, Uzair Ahmed; Zhang, Weihong

    2013-10-01

    The use of composite pressure vessels particularly in the aerospace industry is escalating rapidly because of their superiority in directional strength and colossal weight advantage. The present work elucidates the procedure to optimize the lay-up for composite pressure vessel using finite element analysis and calculate the relative weight saving compared with the reference metallic pressure vessel. The determination of proper fiber orientation and laminate thickness is very important to decrease manufacturing difficulties and increase structural efficiency. In the present work different lay-up sequences for laminates including, cross-ply [ 0 m /90 n ] s , angle-ply [ ±θ] ns , [ 90/±θ] ns and [ 0/±θ] ns , are analyzed. The lay-up sequence, orientation and laminate thickness (number of layers) are optimized for three candidate composite materials S-glass/epoxy, Kevlar/epoxy and Carbon/epoxy. Finite element analysis of composite pressure vessel is performed by using commercial finite element code ANSYS and utilizing the capabilities of ANSYS Parametric Design Language and Design Optimization module to automate the process of optimization. For verification, a code is developed in MATLAB based on classical lamination theory; incorporating Tsai-Wu failure criterion for first-ply failure (FPF). The results of the MATLAB code shows its effectiveness in theoretical prediction of first-ply failure strengths of laminated composite pressure vessels and close agreement with the FEA results. The optimization results shows that for all the composite material systems considered, the angle-ply [ ±θ] ns is the optimum lay-up. For given fixed ply thickness the total thickness of laminate is obtained resulting in factor of safety slightly higher than two. Both Carbon/epoxy and Kevlar/Epoxy resulted in approximately same laminate thickness and considerable percentage of weight saving, but S-glass/epoxy resulted in weight increment.

  9. Fatigue crack growth control in pressure vessel using single peak overload and load hold at elevated temperatures

    International Nuclear Information System (INIS)

    Chen, Ling; Nakamura, Haruo; Kobayashi, Hideo

    1995-01-01

    Pressure vessels are periodically subjected to a proof test to assure structural integrity. During that period, it may be possible to control the fatigue crack growth by simultaneous heating globally or locally around a crack tip. From this viewpoint, retardation behavior of a fatigue crack due to a single peak overload at a high temperature (∼300degC) was examined in an A 533 B-1 steel. Parameters investigated were overload ratio, applied temperature, hold time at higher temperature, and unloading temperature of overload. The higher the applied temperature, the greater the effect of retardation especially when unloading of the overload is done at lower temperatures. Also, the hold time at maximum load promotes oxide-film-induced crack closure. To evaluate those effects, a simple model is proposed based on the Dugdale model. (author)

  10. 46 CFR 52.01-2 - Adoption of section I of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section I of the ASME Boiler and Pressure... and Pressure Vessel Code. (a) Main power boilers and auxiliary boilers shall be designed, constructed, inspected, tested, and stamped in accordance with section I of the ASME Boiler and Pressure Vessel Code...

  11. 46 CFR 53.01-3 - Adoption of section IV of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section IV of the ASME Boiler and Pressure... Boiler and Pressure Vessel Code. (a) Heating boilers shall be designed, constructed, inspected, tested, and stamped in accordance with section IV of the ASME Boiler and Pressure Vessel Code (incorporated by...

  12. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    International Nuclear Information System (INIS)

    Roche, R.; Gaudez, J.C.

    1964-01-01

    In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

  13. Preventive protection device and method for bottom of reactor pressure vessel

    International Nuclear Information System (INIS)

    Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

    1995-01-01

    In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

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

    International Nuclear Information System (INIS)

    Yahr, G.T.

    1995-01-01

    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 √m (23 ksi √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

  15. Structure mechanical analysis of prestressed cast-steel pressure vessels with the finite-element-method

    International Nuclear Information System (INIS)

    Edalat, B.

    1981-08-01

    The analytical pressure analysis is performed for a vessel with solid bottom and top. The basis of the Finite-Element-Method (FEM) and the criteria for the choice of a suitable element type for use in the computer model was investigated. To investigate the exactness of the FE-program a comparison between the analytical solution and the pressure claculated by FEM at a cylindrical vessel was made. For pressure analyses at the test vessel built of steel sections four different computer models (after FEM) were developed. The pressure analysis of a prestressed cast-steel pressure vessel for the transport and for the storage of burnt HTR fuel elements is performed with the aid of computed models after FEM. The method of developing simple computer models for the prestressed pressure vessel with large dimension is explained with an example. (orig.) [de

  16. Seals for sealing a pressure vessel such as a nuclear reactor vessel or the like

    International Nuclear Information System (INIS)

    Bruns, H.J.; Huelsermann, K.H.

    1975-01-01

    A description is given of seals for sealing a pressure vessel such as a nuclear reactor vessel, steam boiler vessel, or any other vessel which is desirably sealed against pressure of the type including a housing and a housing closure that present opposed vertical sealing surfaces which define the sides of a channel. The seals of the present invention comprise at least one sealing member disposed in the channel, having at least one stop face, a base portion and two shank portions extending from the base portion to form a groove-like recess. The shank portions are provided with sealing surfaces arranged to mate with the opposed vertical pressure vessel sealing surfaces. A shank-spreading wedge element also disposed in the channel has at least one stop face and is engaged in the groove-like recess with the sealing member and wedge element stop face adjacent to each other

  17. Evaluation of Creep-Fatigue Integrity for High Temperature Pressure Vessel in a Sodium Test Loop

    International Nuclear Information System (INIS)

    Lee, Hyeong-Yeon; Lee, Dong-Won

    2014-01-01

    In this study, high temperature integrity evaluation on a pressure vessel of the expansion tank operating at elevated temperature of 510°C in the sodium test facility of the SEFLA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) to be constructed at KAERI has been performed. Evaluations of creep fatigue damage based on a full 3D finite element analyses were conducted for the expansion tank according to the recent elevated temperature design codes of ASME Section III Subsection NH and French RCC-MRx. It was shown that the expansion tank maintains its integrity under the intended creep-fatigue loads. Quantitative code comparisons were conducted for the pressure vessel of austenitic stainless steel 316L

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

    Energy Technology Data Exchange (ETDEWEB)

    Neyra Astudillo, Miriam Rocío, E-mail: neyra@cnea.gov.ar [IT Sabato, Universidad Nacional de San Martín, UNSAM, Av. General Paz 1499, Buenos Aires (Argentina); Universidad Tecnológica Nacional UTN, Regional Delta, Buenos Aires (Argentina); López Pumarega, María Isabel, E-mail: lopezpum@cnea.gov.ar [Comisión Nacional de Energía Atómica, CNEA, Av. General Paz 1499, Buenos Aires (Argentina); Núñez, Nicolás Marcelo, E-mail: nnunez@cnea.gov.ar [Comisión Nacional de Energía Atómica, CNEA, Av. General Paz 1499, Buenos Aires (Argentina); Pochettino, Alberto, E-mail: alberto.poch@gmail.com [Comisión Nacional de Energía Atómica, CNEA, Av. General Paz 1499, Buenos Aires (Argentina); Instituto de Investigación e Ingeniería Ambiental (3iA), Campus Miguelete, UNSAM, Av. 25 de Mayo y Francia, 1650 San Martín Argentina (Argentina); Ruzzante, José, E-mail: ruzzante@gmail.com [Universidad Tecnológica Nacional UTN, Regional Delta, Buenos Aires (Argentina); Universidad Nacional de Tres de Febrero UNTREF, Caseros, Buenos Aires (Argentina); Universidad Nacional de Chilecito, UNdeC, La Rioja (Argentina)

    2017-03-15

    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 - Highlights: • MBN and MAE studied in nuclear power pressure vessel steel. • Comparison with macro graphic study of sulfides and texture analysis of the material. • Connections with magnetic anisotropy, texture and microstructure of material. • Plastic flow direction different from the forging direction indicated.

  19. Considerations for acoustic emission monitoring of spherical Kevlar/epoxy composite pressure vessels

    Science.gov (United States)

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

    1977-01-01

    We are continuing to research the applications of acoustic emission testing for predicting burst pressure of filament-wound Kevlar 49/epoxy pressure vessels. This study has focused on three specific areas. The first area involves development of an experimental technique and the proper instrumentation to measure the energy given off by the acoustic emission transducer per acoustic emission burst. The second area concerns the design of a test fixture in which to mount the composite vessel so that the acoustic emission transducers are held against the outer surface of the composite. Included in this study area is the calibration of the entire test setup including couplant, transducer, electronics, and the instrument measuring the energy per burst. In the third and final area of this study, we consider the number, location, and sensitivity of the acoustic emission transducers used for proof testing composite pressure vessels.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  1. Basic requirements of mechanical properties for nuclear pressure vessel materials in ASME-BPV code

    International Nuclear Information System (INIS)

    Ning Dong; Yao Weida

    2011-01-01

    The four basic aspects of strengths, ductility, toughness and fatigue strengths can be summarized for overall mechanical properties requirements of materials for nuclear pressure-retaining vessels in ASME-BPV code. These mechanical property indexes involve in the factors of melting, manufacture, delivery conditions, check or recheck for mechanical properties and chemical compositions, etc. and relate to degradation and damage accumulation during the use of materials. This paper specifically accounts for the basic requirements and theoretic basis of mechanical properties for nuclear pressure vessel materials in ASME-BPV code and states the internal mutual relationships among the four aspects of mechanical properties. This paper focuses on putting forward at several problems on mechanical properties of materials that shall be concerned about during design and manufacture for nuclear pressure vessels according to ASME-BPV code. (author)

  2. Pressure vessel with improved impact resistance and method of making the same

    Science.gov (United States)

    DeLay, Thomas K. (Inventor); Patterson, James E. (Inventor); Olson, Michael A. (Inventor)

    2010-01-01

    A composite overwrapped pressure vessel is provided which includes a composite overwrapping material including fibers disposed in a resin matrix. At least first and second kinds of fibers are used. These fibers typically have characteristics of high strength and high toughness to provide impact resistance with increased pressure handling capability and low weight. The fibers are applied to form a pressure vessel using wrapping or winding techniques with winding angles varied for specific performance characteristics. The fibers of different kinds are dispersed in a single layer of winding or wound in distinct separate layers. Layers of fabric comprised of such fibers are interspersed between windings for added strength or impact resistance. The weight percentages of the high toughness and high strength materials are varied to provide specified impact resistance characteristics. The resin matrix is formed with prepregnated fibers or through wet winding. The vessels are formed with or without liners.

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

    Science.gov (United States)

    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.

  4. INETEC new system for inspection of PWR reactor pressure vessel head

    International Nuclear Information System (INIS)

    Nadinic, B.; Postruzin, Z.

    2004-01-01

    INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

  5. Self-sensing of carbon nanofiber concrete columns subjected to reversed cyclic loading

    Science.gov (United States)

    Howser, R. N.; Dhonde, H. B.; Mo, Y. L.

    2011-08-01

    Civil infrastructures are generally a country's most expensive investment, and concrete is the most widely used material in the construction of civil infrastructures. During a structure's service life, concrete ages and deteriorates, leading to substantial loss of structural integrity and potentially resulting in catastrophic disasters such as highway bridge collapses. A solution for preventing such occurrences is the use of structural health monitoring (SHM) technology for concrete structures containing carbon nanofibers (CNF). CNF concrete has many structural benefits. CNF restricts the growth of nanocracks in addition to yielding higher strength and ductility. Additionally, test results indicate a relationship between electrical resistance and concrete strain, which can be well utilized for SHM. A series of reinforced concrete (RC) columns were built and tested under a reversed cyclic loading using CNF as a SHM device. The SHM device detected and assessed the level of damage in the RC columns, providing a real-time health monitoring system for the structure's overall integrity.

  6. Self-sensing of carbon nanofiber concrete columns subjected to reversed cyclic loading

    International Nuclear Information System (INIS)

    Howser, R N; Dhonde, H B; Mo, Y L

    2011-01-01

    Civil infrastructures are generally a country's most expensive investment, and concrete is the most widely used material in the construction of civil infrastructures. During a structure's service life, concrete ages and deteriorates, leading to substantial loss of structural integrity and potentially resulting in catastrophic disasters such as highway bridge collapses. A solution for preventing such occurrences is the use of structural health monitoring (SHM) technology for concrete structures containing carbon nanofibers (CNF). CNF concrete has many structural benefits. CNF restricts the growth of nanocracks in addition to yielding higher strength and ductility. Additionally, test results indicate a relationship between electrical resistance and concrete strain, which can be well utilized for SHM. A series of reinforced concrete (RC) columns were built and tested under a reversed cyclic loading using CNF as a SHM device. The SHM device detected and assessed the level of damage in the RC columns, providing a real-time health monitoring system for the structure's overall integrity

  7. Efficiency of fiber reinforced concrete application in structures subjected to dynamic effects

    Directory of Open Access Journals (Sweden)

    Morozov Valeriy Ivanovich

    2014-03-01

    Full Text Available Fiber reinforced concretes possess high strength under dynamic loadings, which include impact loads, thanks to their high structural viscosity. This is the reason for using them in difficult operating conditions, where increasing the performance characteristics and the structure durability is of prime importance, and the issues of the cost become less significant. Applying methods of disperse reinforcement is most challenging in case of subtle high-porous materials on mineral binders, for example foamed concrete. At the same time, the experiments conducted in Russia and abroad show, that also in other cases the concrete strength resistance several times increases as a result of disperse reinforcement. This doesn't depend on average density of the concrete and type of fiber used. In the article the fibre reinforced concrete impact resistance is analysed. Recommendations are given in regard to fibre concrete application in manufacture of monolithic floor units for industrial buildings and precast piles.

  8. Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

    1980-01-01

    The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

  9. Analysis of mechanical property data obtained from nuclear pressure vessel surveillance capsules

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1977-01-01

    A typical pressure vessel surveillance capsule examination program provides mechanical property data from tensile, Charpy V-notch impact, and, in some cases, fracture mechanics specimens. This data must be analyzed in conjunction with the unirradiated baseline mechanical property data to determine the effect of irradiation on the mechanical properties. In the case of Charpy impact specimens, for example, irradiation typically causes an increase in the transition temperature, and a decrease in the upper shelf energy level. The results of the Charpy impact and other mechanical specimen tests must be evaluated to determine if property changes are occurring in the manner expected when the reactor was put into service. The large amount of data obtained from surveillance capsule examinations in recent years enables one to make fairly good predictions. After the changes in the mechanical properties of specimens from a particular surveillance capsule have been experimentally determined and evaluated, they must be related to the reactor pressure vessel. This requires a knowledge of the neutron fluence of the surveillance capsule, and the ratio of the surveillance capsule fluence to the pressure vessel wall fluence. This ratio is frequently specified by the reactor manufacturer, or can be calculated from a knowledge of the geometry and materials of the reactor components inside the pressure vessel. A knowledge of the exact neutron fluence of the capsule specimens and the capsule to vessel wall neutron fluence ratio is of great importance, since inaccuracies in these numbers cause just as serious a problem as inaccuracies in the mechanical property determinations. A further area causing analysis difficulties is problems encountered in recent capsule programs relating to capsule design, construction, operation, and dismantling. (author)

  10. Improving the procedure of manufacturing the reactor pressure vessel head for WWER-1000

    International Nuclear Information System (INIS)

    Shutkov, G.A.; Zorin, V.G.; Kryuger, E.A.

    1985-01-01

    A procedure is suggested for manufacturing the reactor pressure vessel heat for the WWER-1000 by means of stamping seamless - forged blakds of he 15Kh2NMFA steel. As a result of the new manufacturing procedure the time adn labour input for the manufacturing, metal consumption are reduced, the product durability is increased due to elimination of welded joints. Mechanical properties of the items completely meet the requirements

  11. Reactor pressure vessel steels[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-07-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use.

  12. Pressure vessel sliding support unit and system using the sliding support unit

    Science.gov (United States)

    Breach, Michael R.; Keck, David J.; Deaver, Gerald A.

    2013-01-15

    Provided is a sliding support and a system using the sliding support unit. The sliding support unit may include a fulcrum capture configured to attach to a support flange, a fulcrum support configured to attach to the fulcrum capture, and a baseplate block configured to support the fulcrum support. The system using the sliding support unit may include a pressure vessel, a pedestal bracket, and a plurality of sliding support units.

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

    Science.gov (United States)

    Greulich, Owen Rudolf

    2017-01-01

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

  14. Lateral expansion and impact toughness correlation of VVER-1000 reactor pressure vessel materials

    Directory of Open Access Journals (Sweden)

    O. V. Trygubenko

    2014-12-01

    Full Text Available Impact toughness and lateral expansion of the irradiated surveillance specimens for VVER-1000 reactor pres-sure vessel have been defined using Charpy impact test. The analysis of experimental data has revealed a linear correlation of these characteristics. It was shown that a slope of the regression line is practically unchanged due to irradiation. Using the upper shelf energy test data it was also demonstrated the lateral expansion and impact toughness decrease simultaneously under the neutron irradiation.

  15. Use of automation and mechanization elements in welding and surfacing nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Bartak, J.; Elckner, J.

    1986-01-01

    The problems are discussed of automation and mechanization of individual operations in the production cycle of pressure vessels whose manufacture cannot for its great exactingness be automated as a whole. Examples are given of workplaces and single-purpose welding facilities with a high level of automation. The present state of the development and implementation of automation of arc welding is described and further development is indicated of the automation of welding processes in the manufacture of nuclear facilities. (J.C.)

  16. Radiation embrittlement of nuclear reactor pressure vessel steels: An international review (second volume)

    International Nuclear Information System (INIS)

    Steele, L.E.

    1986-01-01

    This book has sixteen (16) peer-reviewed papers divided into four (4) sections that reflect changes in the nuclear power industry occurring since 1981, including escalating capital requirements and a growing worldwide dependence on nuclear power for electricity production. The four (4) sections of this book are: Overview of National Programs; Surveillance and Other Radiation Embrittlement Studies; Pressure Vessel Integrity and Regulatory Considerations; and Mechanisms of Irradiation Embrittlement

  17. Powering Small Satellites with Advanced NiH2 Dependent Pressure Vessel (DPV) Batteries

    OpenAIRE

    Caldwell, Dwight; Fox, Chris; Miller, Lee

    1996-01-01

    The Dependent Pressure Vessel (DPV) nickel-hydrogen (NiH2) design is being developed by Eagle-Picher Industries, Inc. (EPI), as a spacecraft battery for both large and small, military and commercial satellites. The DPV cell design offers high specific energy, energy density and reduced cost, while retaining the established IPV technology flight heritage and database. This advanced design also offers a more efficient mechanical, electrical and thermal cell and battery configuration and a reduc...

  18. Preliminary investigation of ultrasonic shear wave holography with a view to the inspection of pressure vessels

    International Nuclear Information System (INIS)

    Aldridge, E.E.; Clare, A.B.; Shepherd, D.A.

    1975-01-01

    The manner in which holography would fit into the general scheme of pressure vessel inspection is discussed. Compared to conventional A, B and C presentations holography requires a different processing of the ultrasonic signal and a mechanical scan which may be more demanding than that normally provided for a C display. Preliminary results are presented of the examination of artificial defects in steel plate using shear wave holography. (author)

  19. Influence of pre/in-service inspections and tests on the reliability of reactor pressure vessels

    International Nuclear Information System (INIS)

    Wellein, R.

    1982-01-01

    The influence of the following actions on the probability of brittle failure of the reactor pressure vessels will be estimated by probabilistic fracture mechanics: ultrasonic inspection of the welds; hydro-test of the vessel; and crack growth by normal, upset and test conditions. Taking into account that the in-service inspections and tests are done at short intervals the reliability can be shown to be extremely high. (orig.)

  20. Evaluation of fatigue damage of pressure vessel materials by observation of microstructures

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1994-01-01

    As the important factor as the secular change mode of pressure vessel materials, there is fatigue damage. In USA, there is the move to use LWRs by extending their life, and it becomes necessary to show the soundness of the structures of machinery and equipment for long period. For exactly evaluating the soundness of the structures of machinery and equipment, it is important to clarify the degree of secular deterioration of the materials. In this report, by limiting to the fatigue damage of LWR pressure vessel steel, the method of grasping the change of microstructure and the method of estimating the degree of fatigue damage from the change of microstructure are shown. The change of microstructure arising in materials due to fatigue advances in the following steps, namely, the multiplication of dislocations, the tangling of dislocations, the formation of cell structure, the turning of cells, the formation of microcracks, the growth of cracks and fracture. In the case of pressure vessel steel, due to the quenching and tempering, the cell structure is formed from the beginning, and the advance of fatigue is recognized as the increase of the turning angle of cell structures. The detection of fatigue damage by microstructure is reported. (K.I.)

  1. Application of an ultrasonic data recording and processing system to reactor pressure vessel examination

    International Nuclear Information System (INIS)

    Buxbaum, S.R.; Pond, R.B. Jr; Willetts, A.J.

    1988-01-01

    Pre-service and periodic in-service nondestructive examinations of nuclear reactor pressure vessels in the USA are mandated by law and are performed in accordance with Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code and Nuclear Regulatory Commission (NRC) requirements. Although acceptable to Code and regulatory requirements, conventional examination procedures and techniques may not meet utility needs, such as providing an adequate technical basis for plant license extension arguments and responding to concerns for nuclear plant safety. A multichannel Ultrasonic Data Recording and Processing System (UDRPS) was developed to address these concerns and was applied to the inspection of reactor pressure vessels in parallel with the Code-required inspection. The inspection requirements and the UDRPS and Code-required conventional NDE data acquisition and analysis are described. Results from the Calvert Cliffs RPV inspections demonstrated that the UDRPS examination was more sensitive than the Code-required examination, and caused only minimal impact on the Code-required examination and overall inspection schedule. (author)

  2. Reactor pressure vessel embrittlement of NPP borssele: Design lifetime and lifetime extension

    International Nuclear Information System (INIS)

    Blom, F.J.

    2007-01-01

    Embrittlement of the reactor pressure vessel of the Borssele nuclear power plant has been investigated taking account of the design lifetime of 40 years and considering 20 years subsequent lifetime extension. The paper presents the current licensing status based on considerations of material test data and of US nuclear regulatory standards. Embrittlement status is also evaluated against German and French nuclear safety standards. Results from previous fracture toughness and Charpy tests are investigated by means of the Master curve toughness transition approach. Finally, state of the art insights are investigated by means of literature research. Regarding the embrittlement status of the reactor pressure vessel of Borssele nuclear power plant it is concluded that there is a profound basis for the current license up to the original end of the design life in 2013. The embrittlement temperature changes only slightly with respect to the acceptance criterion adopted postulating further operation up to 2033. Continued safe operation and further lifetime extension are therefore not restricted by reactor pressure vessel embrittlement

  3. Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

    International Nuclear Information System (INIS)

    Mishra, J.; Balasubramaniyan, V.

    2016-01-01

    A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

  4. New approaches for evaluation of brittle strength of reactor pressure vessels

    International Nuclear Information System (INIS)

    Margolin, B.; Karzov, G.; Kostylev, V.; Gulenko, A.; Rivkin, E.

    2003-01-01

    Based on the Master curve conception, condition of brittle strength is formulated for heterogeneous distribution of stress intensity factor along crack front and non-monotonic, non-isothermic loading. This formulation includes the elaborated procedure for taking into account the effects of shallow cracks and biaxial loading on fracture toughness. It is concluded as follows: (1) A condition of brittle strength is formulated for reactor pressure vessel with crack-like flaw in probabilistic statement. As the condition of brittle strength it is taken condition P f f (bar) where P f is fracture probability, and P f (bar) is a given level of fracture probability. Formulation of this condition is based on the weakest link model and takes into account a variation of the stress intensity factor K I and K IC along the crack front. (2) Dependencies are proposed which allow to take into account the shallow crack effect and the biaxial loading effect on fracture toughness of reactor pressure vessel steels. (3) Using approaches presented in the present paper allows one to decrease conservatism and to increase adequacy of evaluations of brittle strength of reactor pressure vessels. Now these approaches have been included in Russian Standard on evaluation of brittle fracture of RPV of WWER type

  5. Stress corrosion cracking of nuclear reactor pressure vessel and piping steels

    International Nuclear Information System (INIS)

    Speidel, M.O.; Magdowski, R.M.

    1988-01-01

    This paper presents an extensive investigation of stress corrosion cracking of nuclear reactor pressure vessel and piping steels exposed to hot water. Experimental fracture mechanics results are compared with data from the literature and other laboratories. Thus a comprehensive overview of the present knowledge concerning stress corrosion crack growth rates is provided. Several sets of data confirm that 'fast' stress corrosion cracks with growth rates between 10 -8 and 10 -7 m/s and threshold stress intensities around 20 MN m -3/2 can occur under certain conditions. However, it appears possible that specific environmental, mechanical and metallurgical conditions which may prevail in reactors can result in significantly lower stress corrosion crack growth rates. The presently known stress corrosion crack growth rate versus stress intensity curves are discussed with emphasis on their usefulness in establishing safety margins against stress corrosion cracking of components in service. Further substantial research efforts would be helpful to provide a data base which permits well founded predictions as to how stress corrosion cracking in pressure vessels and piping can be reliably excluded or tolerated. It is emphasized, however, that the nucleation of stress corrosion cracks (as opposed to their growth) is difficult and may contribute substantially to the stress corrosion free service behaviour of the overwhelming majority of pressure vessels and pipes. (author)

  6. Analysis and evaluation system for elevated temperature design of pressure vessels

    International Nuclear Information System (INIS)

    Hayakawa, Teiji; Sayawaki, Masaaki; Nishitani, Masahiro; Mii, Tatsuo; Murasawa, Kanji

    1977-01-01

    In pressure vessel technology, intensive efforts have recently been made to develop the elevated temperature design methods. Much of the impetus of these efforts has been provided mainly by the results of the Liquid Metal Fast Breeder Reactor (LMFBR) and more recently, of the High Temperature Gas-cooled Reactor (HTGR) Programs. The pressure vessels and associated components in these new type nuclear power plants must operate for long periods at elevated temperature where creep effects are significant and then must be designed by rigorous analysis for high reliability and safety. To carry out such an elevated temperature designing, numbers of highly developed analysis and evaluation techniques, which are so complicated as to be impossible by manual work, are indispensable. Under these circumstances, the authors have made the following approaches in the study: (1) Study into basic concepts and the associated techniques in elevated temperature design. (2) Systematization (Analysis System) of the procedure for loads and stress analyses. (3) Development of post-processor, ''POST-1592'', for strength evaluation based on ASME Code Case 1592-7. By linking the POST-1592 together with the Analysis System, an analysis and evaluation system is developed for an elevated temperature design of pressure vessels. Consequently, designing of elevated temperature vessels by detailed analysis and evaluation has easily and effectively become feasible by applying this software system. (auth.)

  7. Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

    International Nuclear Information System (INIS)

    Boros, I.; Aszodi, A.

    2003-01-01

    The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

  8. Experimental Study of Confined Low-, Medium- and High-Strength Concrete Subjected to Concentric Compression

    Directory of Open Access Journals (Sweden)

    Antonius

    2012-11-01

    Full Text Available An experimental study of 23 low-, medium- and high-strength concrete columns is presented in this paper. Square-confined concrete columns without longitudinal reinforcement were designed, and tested under concentric axial compression. The columns were made of concrete with a compressive strength ranging between 30 MPa and 70 MPa. The test parameters in the study are concrete compressive strengths and confining steel properties, i.e. spacing, volumetric ratios and configurations. The effects of these parameters on the strength and ductility of square-confined concrete were evaluated. Of the specimens tested in this study, the columns made with higher-strength concrete produced less strength enhancement and ductility than those with lower-strength concrete. The steel configurations were found to have an important role in governing the strength and ductility of the confined high-strength concrete. Moreover, several models of strength enhancement for confined concrete available in the literature turned out to be quite accurate in predicting the experimental results.

  9. A software prototype for assessing the reliability of a concrete bridge superstructure subjected to chloride-induced reinforcement corrosion

    DEFF Research Database (Denmark)

    Schneider, Ronald; Thöns, Sebastian; Fischer, Johannes

    2014-01-01

    A software prototype is developed for assessing and updating the reliability of single-cell prestressed concrete box girders subjected to chloride-induced reinforcement corrosion. The underlying system model consists of two integrated sub-models: a condition model for predicting the deterioration...

  10. Matching the results of a theoretical model with failure rates obtained from a population of non-nuclear pressure vessels

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1982-02-01

    Failure rates for non-nuclear pressure vessel populations are often regarded as showing a decrease with time. Empirical evidence can be cited which supports this view. On the other hand theoretical predictions of PWR type reactor pressure vessel failure rates have shown an increasing failure rate with time. It is shown that these two situations are not necessarily incompatible. If adjustments are made to the input data of the theoretical model to treat a non-nuclear pressure vessel population, the model can produce a failure rate which decreases with time. These adjustments are explained and the results obtained are shown. (author)

  11. Contribution for the improvement of pressurized thermal shock assessment methodologies in PWR pressure vessels

    International Nuclear Information System (INIS)

    Gomes, Paulo de Tarso Vida

    2005-01-01

    The structural integrity assessment of nuclear reactor pressure vessel, concerned to Pressurized Thermal Shock (PTS) accidents, became a necessity and has been investigated since the eighty's. The recognition of the importance of PTS assessment has led the international nuclear technology community to devote a considerable research effort directed to the complete integrity assessment process of the Reactor Pressure Vessels (VPR). Researchers in Europe, Japan and U.S.A. have concentrated efforts in the VPR structural and fracture analysis, conducting experiments to best understand how specific factors act on the behavior of discontinuities, under PTS loading conditions. The main goal of this work is to study de structural behavior of an 'in scale' PWR nuclear reactor pressure vessel model, containing actual discontinuities, under loading conditions generated by a pressurized thermal shock. To construct the pressure vessel model utilized in this research, the approach developed by Barroso (1995) and based on likelihood studies, related to thermal-hydraulic behavior during the PTS was employed. To achieve the objective of this research, a new methodology to generate cracks, with known geometry and localization in the vessel model wall was developed. Additionally, an hydraulic circuit, able to flood the vessel model, heated to 300 deg C, with 10 m 3 of water at 8 deg C, in 170 seconds, was built. Thermo-hydraulic calculations using RELAP5/M0D 3.2.2γ computational code were done, to estimate the temperature profiles during the cooling time. The resulting data subsidized the thermo-structural calculations that were accomplished using ANSYS 7.01 computational code, for both 2D and 3D models. So, the stress profiles obtained with these calculations were associated with fracture mechanics concepts, to assess the crack growth behavior in the VPR model wall. After the PTS test, the VPR model was submitted to destructive and non-destructive inspections. The results

  12. Razpoke odkrite pri medobratovalnem pregledu reaktorske posode in ocena njihove sprejemljivosti: The flaws detected by in-service inspection in reactor pressure vessel and assessment of their acceptability:

    OpenAIRE

    Korošec, Darko; Vojvodič-Tuma, Jelena

    1998-01-01

    By the ASME-Pressure Vessel and Boilers Code, Sec. XI and its specific requirements periodic nondestructive examination of the reactor pressure vessel is required. Reactor pressure vessel is one of the most important pressure boundaries in the pressurized water reactor type. In the present paper the ASME XI code requirements about proper flow determination regarding their size are described. The nondestructive examination on reactor pressure vessel is performed mainly using ultrasonic techniq...

  13. Strain Behavior of Concrete Panels Subjected to Different Nose Shapes of Projectile Impact.

    Science.gov (United States)

    Lee, Sangkyu; Kim, Gyuyong; Kim, Hongseop; Son, Minjae; Choe, Gyeongcheol; Nam, Jeongsoo

    2018-03-09

    This study evaluates the fracture properties and rear-face strain distribution of nonreinforced and hooked steel fiber-reinforced concrete panels penetrated by projectiles of three different nose shapes: sharp, hemispherical, and flat. The sharp projectile nose resulted in a deeper penetration because of the concentration of the impact force. Conversely, the flat projectile nose resulted in shallower penetrations. The penetration based on different projectile nose shapes is directly related to the impact force transmitted to the rear face. Scabbing can be more accurately predicted by the tensile strain on the rear face of concrete due to the projectile nose shape. The tensile strain on the rear face of the concrete was reduced by the hooked steel fiber reinforcement because the hooked steel fiber absorbed some of the impact stress transmitted to the rear face of the concrete. Consequently, the strain behavior on the rear face of concrete according to the projectile nose shape was confirmed.

  14. Properties and Behavior of Geopolymer Concrete Subjected to Explosive Air Blast Loading: A Review

    Directory of Open Access Journals (Sweden)

    Mohd Mortar Nurul Aida

    2017-01-01

    Full Text Available The severe damage to civilian buildings, public area, jet aircraft impact and defense target under explosive blast loading can cause a huge property loss. Most of researcher discusses the topics on design the concrete material model to sustain againts the explosive detonation. The implementation of modern reinforcement steels and fibres in ordinary Portland cement (OPC concrete matrix can reduce the extreme loading effects. However, most researchers have proved that geopolymer concrete (GPC has better mechanical properties towards high performance concrete, compared to OPC. GPC has the high early compressive strength and high ability to resist the thermal energy from explosive detonation. In addition, OPC production is less environmental friendly than geopolymer cement. Geopolymer used can lead to environmental protection besides being improved in mechanical properties. Thus, this paper highlighted on an experimental, numerical and the analytical studies cause of the explosive detonation impact to concrete structures.

  15. Study on steel plate reinforced concrete panels subjected to cyclic in-plane shear

    International Nuclear Information System (INIS)

    Ozaki, Masahiko; Akita, Shodo; Osuga, Hiroshi; Nakayama, Tatsuo; Adachi, Naoyuki

    2004-01-01

    This paper describes the derivation of the equation for evaluating the strength of steel plate reinforced concrete structure (SC) and the experimental results of SC panels subjected to in-plane shear. Two experimental research programs were carried out. One was the experimental study in which the influence of the axial force and the partitioning web were investigated, another was that in which the influence of the opening was investigated. In the former program, nine specimens were loaded in cyclic in-plane shear. The test parameters were the thickness of the surface steel plate, the effects of the partitioning web and the axial force. The experimental results were compared with the calculated results, and good agreement between the calculated results and the experimental results was shown. In the later programs, six specimens having an opening were loaded in cyclic in-plane shear, and were compared with the results of the specimen without opening. FEM analysis was used to supplement experimental data. Finally, we proposed the equation to calculate the reduction ratio from the opening for design

  16. Composite Behavior of Insulated Concrete Sandwich Wall Panels Subjected to Wind Pressure and Suction.

    Science.gov (United States)

    Choi, Insub; Kim, JunHee; Kim, Ho-Ryong

    2015-03-19

    A full-scale experimental test was conducted to analyze the composite behavior of insulated concrete sandwich wall panels (ICSWPs) subjected to wind pressure and suction. The experimental program was composed of three groups of ICSWP specimens, each with a different type of insulation and number of glass-fiber-reinforced polymer (GFRP) shear grids. The degree of composite action of each specimen was analyzed according to the load direction, type of the insulation, and number of GFRP shear grids by comparing the theoretical and experimental values. The failure modes of the ICSWPs were compared to investigate the effect of bonds according to the load direction and type of insulation. Bonds based on insulation absorptiveness were effective to result in the composite behavior of ICSWP under positive loading tests only, while bonds based on insulation surface roughness were effective under both positive and negative loading tests. Therefore, the composite behavior based on surface roughness can be applied to the calculation of the design strength of ICSWPs with continuous GFRP shear connectors.

  17. Composite Behavior of Insulated Concrete Sandwich Wall Panels Subjected to Wind Pressure and Suction

    Directory of Open Access Journals (Sweden)

    Insub Choi

    2015-03-01

    Full Text Available A full-scale experimental test was conducted to analyze the composite behavior of insulated concrete sandwich wall panels (ICSWPs subjected to wind pressure and suction. The experimental program was composed of three groups of ICSWP specimens, each with a different type of insulation and number of glass-fiber-reinforced polymer (GFRP shear grids. The degree of composite action of each specimen was analyzed according to the load direction, type of the insulation, and number of GFRP shear grids by comparing the theoretical and experimental values. The failure modes of the ICSWPs were compared to investigate the effect of bonds according to the load direction and type of insulation. Bonds based on insulation absorptiveness were effective to result in the composite behavior of ICSWP under positive loading tests only, while bonds based on insulation surface roughness were effective under both positive and negative loading tests. Therefore, the composite behavior based on surface roughness can be applied to the calculation of the design strength of ICSWPs with continuous GFRP shear connectors.

  18. Fracture Failure of Reinforced Concrete Slabs Subjected to Blast Loading Using the Combined Finite-Discrete Element Method

    Directory of Open Access Journals (Sweden)

    Z. M. Jaini

    Full Text Available Abstract Numerical modeling of fracture failure is challenging due to various issues in the constitutive law and the transition of continuum to discrete bodies. Therefore, this study presents the application of the combined finite-discrete element method to investigate the fracture failure of reinforced concrete slabs subjected to blast loading. In numerical modeling, the interaction of non-uniform blast loading on the concrete slab was modeled using the incorporation of the finite element method with a crack rotating approach and the discrete element method to model crack, fracture onset and its post-failures. A time varying pressure-time history based on the mapping method was adopted to define blast loading. The Mohr-Coulomb with Rankine cut-off and von-Mises criteria were applied for concrete and steel reinforcement respectively. The results of scabbing, spalling and fracture show a reliable prediction of damage and fracture.

  19. Ultimate analysis of a 1/4-scale prestressed concrete containment vessel model subject to internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jung; Choun, Young Sun; Lee, Sang Jin; Choi, In Kil; Kim, Hyun Ah [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    The research on the investigation of ultimate capacity and integrity of the containment structures has been internationally performed since the late 1980's. However, it is almost impossible to predict the behavior and ultimate capacity of concrete structures with enough accuracy, because of the uncertainties in material properties of concrete. Especially it is a difficult task to predict the response of containment structures with numerical methods since the complex behaviors of concrete appear with crack formation. The objectives of this research are to establish and develop nonlinear analysis procedures for ultimate capacity of prestressed concrete containment structure subject to internal pressure. In this research 20 and 3D numerical analysis procedures are accomplished and fully evaluated by the test result of 1/4-scale model of a prestressed concrete containment that was tested by SNL. The computer program ABAQUS was used to analyze the 1/4-scale model. There is the limitation in the estimation of nonlinear response of containment with 2D analysis since it simple and doesn't consider penetrations although it has been widely used. Therefore in this research 3D FE analysis considering discontinuity was performed to estimate the response of containment together with 2D FE analysis. And the results of analysis were compared with the results of the pretest Round Robin Analysis of the PCCV model to examine the validity of analytical methods. 14 refs., 40 figs., 22 tabs. (Author)

  20. Proceedings of the 3rd international conference on heat exchangers, boilers and pressure vessels (HEB-97). Vol.1 (Research Papers)

    International Nuclear Information System (INIS)

    1997-04-01

    This conference was held on 5-6 Apr 1997 in Alexandria. the specialists discussed heat exchangers, boilers and pressure vessels. more than 200 papers were presented in the meetings. it contains of data, figures and tables

  1. General Description of the Mechanic Design of the Pressure Vessel and the Internal Mechanical Component of the CAREM Reactor

    International Nuclear Information System (INIS)

    Diez, F.; Horro, R.

    2000-01-01

    This paper presents a brief description of the CAREM reactor pressure vessel and its main internal mechanical components and summarizes the functional requirements and approaches applied for their design, together with a review of the normative applicable in each case

  2. Dosimetry of VVER-1000 Reactor Pressure Vessel and Surveillance Specimens as a part of PLiM at Ukrainian NPPs

    International Nuclear Information System (INIS)

    Bukanov, V.N.; Diemokhin, V.L.; Grytsenko, O.V.; Ilkovych, V.V.; Pugach, A.M.; Pugach, S.M.; Vasylieva, O.G.; Vyshnevskyi, I.M.; Kasatkin, O.G.

    2012-01-01

    A regular surveillance program for VVER-1000 and its shortages are described. The Methodology for determination of neutron flux functionals on surveillance specimens of VVER-1000 pressure vessel is presented. The radiation exposure monitoring system for VVER-1000 pressure vessel is described. The main principles of an additional surveillance program for VVER-1000 are presented. The Dosimetry Experiment, which is already carrying out at Unit 3 of Rivne NPP, is described. (author)

  3. Guidelines for Application of the Master Curve Approach to Reactor Pressure Vessel Integrity in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lyssakov, V.N.; Kang, K.S.

    2005-01-01

    These guidelines have been developed under an International Atomic Energy Agency (IAEA) Co-ordinated Research Project (CRP) titled ''Surveillance Programme Results Application to Reactor Pressure Vessel Integrity Assessment.'' The IAEA has sponsored a series of five CRPs that have led to a focus on measuring the best irradiation fracture parameters using relatively small test specimens for assuring structural integrity of reactor pressure vessel (RPV) materials in Nuclear Power Plants (NPPs)

  4. Ultimate analysis of PWR prestressed concrete containment subjected to internal pressure

    International Nuclear Information System (INIS)

    Hu, H.-T.; Lin, Y.-H.

    2006-01-01

    Numerical analyses are carried out by using the ABAQUS finite element program to predict the ultimate pressure capacity and the failure mode of the PWR prestressed concrete containment at Maanshan nuclear power plant. Material nonlinearity such as concrete cracking, tension stiffening, shear retention, concrete plasticity, yielding of prestressing tendon, yielding of steel reinforcing bar and degradation of material properties due to high temperature are all simulated with proper constitutive models. Geometric nonlinearity due to finite deformation has also been considered. The results of the analysis show that when the prestressed concrete containment fails, extensive cracks take place at the apex of the dome, the junction of the dome and cylinder, and the bottom of the cylinder connecting to the base slab. In addition, the ultimate pressure capacity of the containment is higher than the design pressure by 86%

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

    Science.gov (United States)

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

    2007-01-01

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

  6. Tall concrete buildings subject to vertically moving fires: A case study approach

    OpenAIRE

    Fletcher, Ian A

    2009-01-01

    Fire in buildings can have a severe impact in terms of both human safety and potential economic loss. This is especially true in the case of fires of such severity that the building structure is damaged. Concrete buildings are traditionally regarded as safe in a fire situation as concrete is non-flammable and exhibits highly insulating material properties. The majority of current research relating to the impact of fire on structures examines other forms of construction. Research of concr...

  7. Homogenised constitutive model dedicated to reinforced concrete plates subjected to seismic solicitations

    International Nuclear Information System (INIS)

    Combescure, Christelle

    2013-01-01

    Safety reassessments are periodically performed on the EDF nuclear power plants and the recent seismic reassessments leaded to the necessity of taking into account the non-linear behaviour of materials when modeling and simulating industrial structures of these power plants under seismic solicitations. A large proportion of these infrastructures is composed of reinforced concrete buildings, including reinforced concrete slabs and walls, and literature seems to be poor on plate modeling dedicated to seismic applications for this material. As for the few existing models dedicated to these specific applications, they present either a lack of dissipation energy in the material behaviour, or no micromechanical approach that justifies the parameters needed to properly describe the model. In order to provide a constitutive model which better represents the reinforced concrete plate behaviour under seismic loadings and whose parameters are easier to identify for the civil engineer, a constitutive model dedicated to reinforced concrete plates under seismic solicitations is proposed: the DHRC (Dissipative Homogenised Reinforced Concrete) model. Justified by a periodic homogenisation approach, this model includes two dissipative phenomena: damage of concrete matrix and internal sliding at the interface between steel rebar and surrounding concrete. An original coupling term between damage and sliding, resulting from the homogenisation process, induces a better representation of energy dissipation during the material degradation. The model parameters are identified from the geometric characteristics of the plate and a restricted number of material characteristics, allowing a very simple use of the model. Numerical validations of the DHRC model are presented, showing good agreement with experimental behaviour. A one dimensional simplification of the DHRC model is proposed, allowing the representation of reinforced concrete bars and simplified models of rods and wire mesh

  8. The significance of cladding material on the integrity of nuclear pressure vessels with cracks

    International Nuclear Information System (INIS)

    Sattari-Far, Iradj.

    1989-05-01

    The significance of the austenitic cladding layer is reviewed in this literature study. The cladding induced stresses are generally not considered when evaluating the severity of flaws in reactor pressure vessels. It has been shown that this emission may be misleading. The necessity to consider the cladding induced stresses is also emphasized in the latest edition of ASME XI. Contrary to what is commonly assumed, the austenitic cladding displays a charpy V transition region with a low ductility. The interface material (HAZ) is the most influenced region by irradiation, and a transition shift of over 100 degree C may be expected. Because of the significant difference in the thermal expansion coefficients of the cladding and the base metal, cladding induced stresses can be set up. Even after PWHT, residual stresses of yield magnitude remain in the cladding and the HAZ at ambient temperature. The cladding induced stresses are temperature dependent and decrease as the temperature increases. The cladding induced stresses have a significant influence on small defects near the inside surface of a pressure vessel. For semielliptical surface cracks, the maximum CTOD-value along the crack front is not found at the deepest point, but in the cladding/base metal interface, having a magnitude three times higher than the value in the deepest point. It implies that this type of crack would propagate along the clad/base material interface. At some point in time, the crack will reach a geometry which may cause such a severe condition at the deepest point that it will start to grow in the depth direction as well. The initiation and growth behaviour of such cracks need to be investigated to be able to assess the significance of cladding on the integrity of nuclear pressure vessels. (author) (50 figs., 33 refs.)

  9. The application of acoustic emission measurements on laboratory testpieces to large scale pressure vessel monitoring

    International Nuclear Information System (INIS)

    Ingham, T.; Dawson, D.G.

    1975-01-01

    A test pressure vessel containing 4 artificial defects was monitored for emission whilst pressure cycling to failure. Testpieces cut from both the failed vessel and from as-rolled plate material were tested in the laboratory. A marked difference in emission characteristics was observed between plate and vessel testpieces. Activity from vessel material was virtually constant after general yield and emission amplitudes were low. Plate testpieces showed maximum activity at general yield and more frequent high amplitude emissions. An attempt has been made to compare the system sensitivities between the pressure vessel test and laboratory tests. In the absence of an absolute calibration device, system sensitivities were estimated using dummy signals generated by the excitation of an emission sensor. The measurements have shown an overall difference in sensitivity between vessel and laboratory tests of approximately 25db. The reduced sensitivity in the vessel test is attributed to a combination of differences in sensors, acoustic couplant, attenuation, and dispersion relative to laboratory tests and the relative significance of these factors is discussed. Signal amplitude analysis of the emissions monitored from laboratory testpieces showed that, whith losses of the order of 25 to 30db, few emissions would be detected from the pressure vessel test. It is concluded that no reliable prediction of acoustic behaviour of a structure may be made from laboratory test unless testpieces of the actual structural material are used. A considerable improvement in detection sensitivity, is also required for reliable detection of defects in low strength ductile materials and an absolute method of system calibration is required between tests

  10. Interpreting ASME limits and philosophy in FEA of pressure vessel parts

    International Nuclear Information System (INIS)

    Bezerra, L.M.; Cruz, J.R.B.; Miranda, C.A.J.; Neto, M.M.

    1995-01-01

    In recent years there has been an effort to interpret finite element (FE) stress results on the light of the ASME B and PV rules and philosophy. Many task groups have issued guidelines on stress linearization and classifications. All those attempts have come up trying to cope modern FE techniques with the rules imposed by the ASME Code. This paper is an independent contribution to the Pressure Vessel Research Council (PVRC) groups which are studying the stress classification and the failure mechanism in a FE framework. This work tries to complement the interesting work by Hollinger and Hechmer presented in the PVP-94 in Minneapolis. In that paper, the authors examined a typical support skirt and showed relations between the skirt collapse load obtained by finite element analysis and the loads allowed from the ASME stress limits. To complement such paper, in the present article, different skirt geometry configurations are analyzed. The configurations here investigated consist of similar support skirts but with different angles of attachments between cylinder and cone parts. It will be possible to observe the influence of the bending stress in the collapse load and its relation to the allowable loads inferred from the ASME limits. A pressure vessel with torispherical head under internal pressure is also examined. Using elastic and limit load FEA, the present paper determines the collapse loads of the configurations. It sets up the relations between these collapse loads, stress categories, and limits dictated by the ASME Code Subsection NB. On the light of NB rules and philosophy, this paper shows how different methods of stress assessment, classification, and limits may influence in the design of a pressure vessel

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  12. Stress corrosion cracking studies of reactor pressure vessel steels. Final report

    International Nuclear Information System (INIS)

    Van Der Sluys, W.A.

    1996-10-01

    The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

  13. Inspection device for external examination of pressure vessels, preferably for ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Figlhuber, D.; Gallwas, J.; Weber, R.; Weber, J.

    1978-01-01

    The inspection device is placed in the annular gap between pressure vessel and biological shield of the BWR. In the annulus there is arranged at least one longitudinal rail which has got vertical guideways. Along it there can be moved on testing paths a manipulator with the ultrasonic search unit. The manipulator drive is outside of the inspection annulus. It is coupled to the manipulator by means of a tension member being guided over a reversing unit mounted at the upper end of the longitudinal rail. As a tension member there may be used a drag chain; the drive and the reversing unit are provided with corresponding chain wheels. (DG) [de

  14. Subcritical crack growth in the ligament between the instrumentation rods of the BBR pressure vessel bottom

    International Nuclear Information System (INIS)

    Marci, G.; Bazant, E.; Kautz, H.R.

    1978-01-01

    A fracture mechanics fatigue analysis is made for an assumed crack emanating from the bore of an instrumentation rod. This assumed crack has partially penetrated the Inconel buttering of the 22 Ni Mo Cr 37 on which the structural Inconel welds are laid. Our analysis shows that the assumed crack could only penetrate 26% of the remaining ligament of the Inconel structural weld as a result of the fatigue crack growth during the entire operating life of the pressure vessel. Therefore a leak caused by a flaw missed during pre-service and in-service non-destructive testing can be excluded. (author)

  15. Irradiation surveillant test of High Flux Engineering Test Reactor (HFETR) pressure vessel

    International Nuclear Information System (INIS)

    Gao Weisen; Cui Yonghai

    1991-08-01

    The result of irradiation surveillant test of HFETR pressure vessel is presented. It shows that after 3.3 x 10 20 n/cm 2 irradiation the irradiation brittleness effect in welding seam (ws) is smaller than in heat-affected zone (HAZ), and the irradiation brittleness effect in HAZ is smaller than in base material (BM). From the view point of irradiation brittleness, under present operation conditions of the HFETR, the safety margin of plasticity and toughness in the WS, BM and HAZ is sufficient within the design life of the reactor

  16. Comparative study on two different seal surface structure for reactor pressure vessel sealing behavior

    International Nuclear Information System (INIS)

    Chen Jun; Xiong Guangming; Deng Xiaoyun

    2014-01-01

    The seal surface structure is very important to reactor pressure vessel (RPV) sealing behavior. In this paper, two 3-D RPV sealing analysis finite models have been established with different seal surface structures, in order to study the influence of two structures. The separation of RPV upper and lower flanges, bolt loads and etc. are obtained, which are used to evaluate the sealing behavior of the RPV. Meanwhile, the comparative analysis of safety margin of two seal surface structural had been done, which provides the theoretical basis for RPV seal structure design optimization. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-01

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

  18. The role of gamma rays and freely-migrating defects in reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    Alexander, D.E.; Rehn, L.E.

    1996-09-01

    Gamma ray effects are often neglected when evaluating reactor pressure vessel (RPV) embrittlement. However, recent analyses indicate that in newer style light water reactors, gamma damage can be a substantial fraction of the total displacement damage experienced by the (RPV); ignoring this damage will lead to errors in embrittlement predictions. Furthermore, gamma rays may be more efficient than fast neutrons at producing freely-migrating defects and as such can impact certain embrittlement mechanisms more effectively than fast neutrons. Consideration of these gamma effects are therefore essential for a more complete understanding of radiation embrittlement

  19. Light water reactor - pressure-vessel neutron spectrometry with solid state track recorders

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Gold, R.; Roberts, J.H.

    1982-01-01

    Solid State Track Recorders have been used to measure absolute fission rates in the Light Water Reactor - Pressure Vessel Simulator at the Oak Ridge Pool Critical Assembly. Measurements were made in the 12/13 configuration with 237 Np, 238 U, and 235 U at the TSF, TSB, PVF, 1/4 T, 1/2 T, and 3/4 T locations. To avoid relative power normalization uncertainties, all locations were measured simultaneously in a single run for each isotope. Comparison of SSTR results are made with the results of measurements performed by NBS and CEN/SCK using benchmark referenced fission chambers and with the trends of theoretical calculations

  20. The Safety Course Design and Operations of Composite Overwrapped Pressure Vessels (COPV)

    Science.gov (United States)

    Saulsberry, Regor; Prosser, William

    2015-01-01

    Following a Commercial Launch Vehicle On-Pad COPV (Composite Overwrapped Pressure Vessels) failure, a request was received by the NESC (NASA Engineering and Safety Center) June 14, 2014. An assessment was approved July 10, 2014, to develop and assess the capability of scanning eddy current (EC) nondestructive evaluation (NDE) methods for mapping thickness and inspection for flaws. Current methods could not identify thickness reduction from necking and critical flaw detection was not possible with conventional dye penetrant (PT) methods, so sensitive EC scanning techniques were needed. Developmental methods existed, but had not been fully developed, nor had the requisite capability assessment (i.e., a POD (Probability of Detection) study) been performed.

  1. Users manual data base MATSURV. Reactor pressure vessel material surveillance data management system

    International Nuclear Information System (INIS)

    Kenworthy, L.D.; Tether, C.D.

    1980-02-01

    This Users Guide to the data management system MATSURV has been prepared to assist the user in all facets of the task of processing data related to reactor pressure vessel materials surveillance; preparation of raw data for input, input of data, modification of existing data, retrieval and display of data, and the creation of data reports. MATSURV is structured upon the System 2000 data base management system which is maintained on the IBM 370/168 computer at National Institutes of Health. An overview of System 2000 is provided

  2. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    International Nuclear Information System (INIS)

    Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.

    2007-01-01

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  3. Strength and reliability analysis of metal-composite overwrapped pressure vessel

    Science.gov (United States)

    Burov, A. E.; Lepikhin, A. M.; Moskvichev, V. V.

    2017-12-01

    Metal-composite overwrapped pressure vessels (MCOPV) have found a wide application in aerospace and aeronautical industries. Such vessels should combine impermeability and high weight efficiency with enhanced long-term safety and durability. To meet these requirements, theoretical and experimental studies on the mechanics of deformation and failure of MCOPV are required. In the paper, the analysis on strength, lifetime and reliability of MCOPV is presented. A high performance of the MCOPV is justified by comparing the calculation results with experiment data obtained on full-scale samples.

  4. Remote disconnect of 87'' diameter Conoseal joint applied to pressure vessel closures

    International Nuclear Information System (INIS)

    Forrest, G.

    The design of the Conoseal joint and its application for the pressure vessel closure are described. The tested joint was made of stainless steel-347. In helium testing, joints of 1/8'' to 87'' in dia showed leaks less than 10 -8 to 2.5x10 -8 cm 3 .s -1 depending on the joint diameter. The total time required for the remote disconnection of the joint segments by an air motor shortens the time needed for closure opening from 60 to 4 hours. Flange and joint packing materials are listed with a view to joint applications at various operating temperatures. (J.B.)

  5. Cohesive modelling of the fracture of a neutron irradiated pressure vessel steel

    International Nuclear Information System (INIS)

    Gomez, F.J.; Valiente, A.; Elices, M.

    2003-01-01

    The cohesive fracture process zone model was used to account for the neutron irradiation embrittlement of a pressure vessel steel. The tensile testing and fracture of axisymmetrically notched round specimens were numerically modelled assuming a rectangular traction separation law and the irradiation effects were introduced by due modification of this law. The results corroborate those of the experiments performed in a previous work. The cohesive strength and the cohesive energy of the cohesive model were not considered as adjusting parameters, but they were determined from the data of conventional tensile tests and fracture toughness tests on the assumption that the failure of the specimens in these tests also follows the cohesive model

  6. Remote through-wall sampling of the Trawsfynydd reactor pressure vessel: an overview

    International Nuclear Information System (INIS)

    Curry, A.; Clayton, R.

    1996-01-01

    This paper summarises the application of robotic equipment for gaining access to and removing through-wall samples from welds of the reactor pressure vessel at Trawsfynydd power station. The environment, which presents hazards due to ionising radiation, radioactive contamination and asbestos bearing materials is described. The means of access, by use of remote vehicles complete with robotic manipulators supported by additional vehicles, is reviewed. The use of Abrasive Water Jet Cutting for sample removal is introduced. The relative advantages and disadvantages of this technique are discussed. (Author)

  7. Rates of chemical reaction and atmospheric heating during core debris expulsion from a pressurized vessel

    International Nuclear Information System (INIS)

    Powers, D.A.; Tarbell, W.W.; Brockman, J.E.; Pilch, M.

    1986-01-01

    Core debris may be expelled from a pressurized reactor vessel during a severe nuclear reactor accident. Experimental studies of core debris expulsion from pressurized vessels have established that the expelled material can be lofted into the atmosphere of the reactor containment as particulate 0.4 to 2 mm in diameter. These particles will vigorously react with steam and oxygen in the containment atmosphere. Data on such reactions during tests with 80 kg of expelled melt will be reported. A model of the reaction rates based on gas phase mass transport will be described and shown to account for atmospheric heating and aerosol generation observed in the tests

  8. LWR (light water reactor) pressure vessel irradiation surveillance dosimetry. Quarterly progress report, January-March 1980

    International Nuclear Information System (INIS)

    Guthrie, G.L.; McElroy, W.N.

    1980-12-01

    The report describes progress made in the Light Water Reactor Pressure Vessel Irradiation Surveillance Dosimetry Program during the reporting period. The primary objective of the multi-laboratory program is to prepare an updated and improved set of dosimetry, damage correlation, and associated reactor analysis ASTM Standards for LWR-PV irradiation surveillance programs. Supporting this objective are a series of analytical and experimental validation and calibration studies in 'Standard, Reference, and Controlled Environment Benchmark Fields', reactor 'Test Regions', and operating power reactor 'Surveillance Positions'

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

    Science.gov (United States)

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

    2007-01-01

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

  10. Kawasaki Steel products for nuclear power plant components and pressure vessels

    International Nuclear Information System (INIS)

    Mori, Hiroshi

    1980-01-01

    Modern steelmakers are facing a problem to assure a high quality of products with minimum expenditure. This is especially true in the case of producing heavy plates and forgings for nuclear power plant components and pressure vessels, one of the end-uses demanding the highest quality requirements existing today. This paper introduces features of KSC products of these steels and describes how they are manufactured, including an outline of major equipment and processes including BOF-LRF, quality assurance system, and some glimpse of efforts for research and development. (author)

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

    Science.gov (United States)

    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.

  12. Strain ageing of nuclear pressure vessel steels A533B and A508 cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Toerroenen, K.

    1978-04-01

    The susceptibility of the reactor pressure vessel steels A533B and A508 cl.2 to strain ageing has been studied using conventional tensile and impact testing of prestrained and aged specimens. The results show a modest susceptibility, seen as an increase in yield strength and Charpy V transition temperatures. The effect of varying alloying additions within the range of normal production was not observed, but the initial mechanical properties clearly affect the strain ageing. The lower the initial yield strength, the higher increase in strength and the lower increase in transition temperature is observed. (author)

  13. Slideline verification for multilayer pressure vessel and piping analysis including tangential motion

    International Nuclear Information System (INIS)

    Van Gulick, L.A.

    1984-01-01

    Nonlinear finite element method (FEM) computer codes with slideline algorithm implementations should be useful for the analysis of prestressed multilayer pressure vessels and piping. This paper presents closed form solutions including the effects of tangential motion useful for verifying slideline implementations for this purpose. The solutions describe stresses and displacements of a long internally pressurized elastic-plastic cylinder initially separated from an elastic outer cylinder by a uniform gap. Comparison of closed form and FEM results evaluates the usefulness of the closed form solution and the validity of the sideline implementation used

  14. Special servicing equipment for reactor pressurized vessel stud hole and stud accessories

    International Nuclear Information System (INIS)

    Li Jianglian

    1999-01-01

    The author briefly introduces the design and manufacture of nuclear island special servicing equipment of Nuclear Power Institute of China. Maintenance process of reactor pressurized vessel (RPV) stud hold and stud accessories the special servicing equipment include RPV flange dummy, closed-circuit television (CCTV) inspection equipment, RPV stud hole expandable comb, RPV stud hole polisher, RPV stud hold thread lubricating equipment, RPV stud hole thread miller and RPV stud hole camera. It is presented how eight kinds of special servicing equipment perform the maintenance process concerning their function, structure, and characteristics, their practical use on site is also introduced

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

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon

    2012-01-01

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

  16. Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

    International Nuclear Information System (INIS)

    Klaar, H.J.

    1975-01-01

    As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

  17. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.

    1978-01-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs

  18. Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

    International Nuclear Information System (INIS)

    James, L.A.

    1977-01-01

    The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

  19. Statistical Analysis of Reactor Pressure Vessel Fluence Calculation Benchmark Data Using Multiple Regression Techniques

    International Nuclear Information System (INIS)

    Carew, John F.; Finch, Stephen J.; Lois, Lambros

    2003-01-01

    The calculated >1-MeV pressure vessel fluence is used to determine the fracture toughness and integrity of the reactor pressure vessel. It is therefore of the utmost importance to ensure that the fluence prediction is accurate and unbiased. In practice, this assurance is provided by comparing the predictions of the calculational methodology with an extensive set of accurate benchmarks. A benchmarking database is used to provide an estimate of the overall average measurement-to-calculation (M/C) bias in the calculations ( ). This average is used as an ad-hoc multiplicative adjustment to the calculations to correct for the observed calculational bias. However, this average only provides a well-defined and valid adjustment of the fluence if the M/C data are homogeneous; i.e., the data are statistically independent and there is no correlation between subsets of M/C data.Typically, the identification of correlations between the errors in the database M/C values is difficult because the correlation is of the same magnitude as the random errors in the M/C data and varies substantially over the database. In this paper, an evaluation of a reactor dosimetry benchmark database is performed to determine the statistical validity of the adjustment to the calculated pressure vessel fluence. Physical mechanisms that could potentially introduce a correlation between the subsets of M/C ratios are identified and included in a multiple regression analysis of the M/C data. Rigorous statistical criteria are used to evaluate the homogeneity of the M/C data and determine the validity of the adjustment.For the database evaluated, the M/C data are found to be strongly correlated with dosimeter response threshold energy and dosimeter location (e.g., cavity versus in-vessel). It is shown that because of the inhomogeneity in the M/C data, for this database, the benchmark data do not provide a valid basis for adjusting the pressure vessel fluence.The statistical criteria and methods employed in

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

  1. Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs) . Volume 2; Appendices

    Science.gov (United States)

    Prosser, William H.

    2014-01-01

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

  2. Irradiation effect on mechanical characteristics of pressure vessel steel, expectation formulae, segregations effect

    International Nuclear Information System (INIS)

    Soulat, P.; Brillaud, C.; Houssin, B.

    1997-01-01

    After a brief indicating of the running conditions of the materials used in PWR pressure vessels, the acquired knowledge concerning the parameters having an effect on embrittlement are reviewed as well as the knowledge on the anticipation, the knowledge on the control program organization and the obtained results. The tensile strength of the steel of welded joints containing segregations zones is determined. The use of the local criteria of the fragile rupture allowing to obtain a statistical description of the material properties is given. (O.M.)

  3. A structural evaluation of the Shippingport reactor pressure vessel for transport impact conditions

    International Nuclear Information System (INIS)

    Witte, M.C.; Chou, C.K.

    1989-01-01

    The Shippingport Atomic Power Station in Shippingport, Pennsylvania, is being decommissioned and dismantled. This government-leased property will be returned, in a radiologically safe condition, to its owner. All radioactive material is being removed from the Shippingport Station and transported for burial to the DOE Hanford Reservation in Richland, Washington. The reactor pressure vessel (RPV) will be transported by barge to Hanford. This paper describes an evaluation of the structural response of the RPV to the normal and accident impact test conditions as required by the Code of Federal Regulations. 3 refs., 5 figs., 3 tabs

  4. Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor

    International Nuclear Information System (INIS)

    Saavedra, Fernando M.

    2001-01-01

    The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es

  5. The behavior of reinforced concrete barriers subjected to the impact of tornado generated deformable missiles

    International Nuclear Information System (INIS)

    McMahon, P.M.; Meyers, B.L.; Buchert, K.P.

    1977-01-01

    The paper presents a general model for the evaluation of local effects damage including, penetration and backface spalling, of reinforced concrete barriers subjected to the impact of deformable tornado generated missiles. The model is based on an approximte force time history which assumes: 1) the initial penetration of the missile occurs without significant deformation of the missile if the strength of the missile is greater than that of the barrier. This portion of the time history is represented by a linear and finite rise time; 2) wrinkling or collapse of the missile occurs when the critical stress of the missile is exceeded. This portion of the time histroy is represented by a constant force-time relationship, although a decreaseing force might be more accurate; 3) while the missile is penetrating and wrinkling both elastic and plastic stress waves are developed in the missile, and compressive and shear stress waves are generated in he target. When the shear waves reach the backface of the slab, doagonal cracks initiating at the end of the penetrating missile are formed. These cracks propagate to the backface reinforcing where splitting cracks are formed. Finally, yield hinge lines form in the plane of reinforcing; 4) repenetration of the missile occurs after the wrinkling has caused a change in missile cross section. This repenetration results from moving the failure cone described in three above, and is also represented by the costant force time history. Using the assumptions, relationships for the penetration depth of the missile the wrinkling length of the missile, the critical missile stress, the time history of the impact and the spalling of the target are developed. (Auth.)

  6. Bond Behavior of Wet-Bonded Carbon Fiber-Reinforced Polymer-Concrete Interface Subjected to Moisture

    Directory of Open Access Journals (Sweden)

    Yiyan Lu

    2018-01-01

    Full Text Available The use of carbon fiber-reinforced polymer (CFRP composite materials to strengthen concrete structures has become popular in coastal regions with high humidity levels. However, many concrete structures in these places remain wet as a result of tides and wave-splashing, so they cannot be completely dried before repair. Therefore, it is vital to investigate the effects of moisture on the initial and long-term bond behavior between CFRP and wet concrete. This research assesses the effects of moisture (i during CFRP application and (ii throughout the service life. Before CFRP bonding, the concrete blocks are preconditioned with a water content of 4.73% (termed “wet-bonding”. Three different epoxy resins are applied to study the bond performance of the CFRP-concrete interface when subjected to moisture (95% relative humidity. A total of 45 double-lap shear specimens were tested at the beginning of exposure and again after 1, 3, 6, and 12 months. All specimens with normal epoxy resins exhibited adhesive failure. The failure mode of specimens with hydrophobic epoxy resin changed from cohesive failure to mixed cohesive/adhesive failure and to adhesive failure according to the duration of exposure. Under moisture conditioning, the maximum shear stress (τmax and corresponding slip (smax of the bond-slip curve first increased and then decreased or fluctuated over time. The same tendency was seen in the ultimate strain transmitted to the CFRP sheet, the interfacial fracture energy (Gf, and the ultimate load (Pu. Analytical models of Gf and Pu for the CFRP-concrete interface under moisture conditioning are presented.

  7. Irradiation embrittlement of reactor pressure vessel steels (Report on analysis of the behaviour of advanced reactor pressure vessel steel under neutron irradiation)

    International Nuclear Information System (INIS)

    Sivaramakrishnan, K.S.; Chatterjee, S.; Anantharaman, S.; Balakrishnan, K.S.; Viswanathan, U.K.

    1984-01-01

    The International Working Group on Reliability of Reactor Pressure Components (IWG-RRPC) sponsored by IAEA launched a Coordinated Research Programme in 1977 to investigate the behaviour of Advanced Pressure Vessel Steels under neutron irradiation. IWG-RRPC included provision of various materials produced by organisation in Federal Republic of Germany, France, Japan and a reference Steel from United States of America to those countries participating in the programme. India joined the International Programme in 1977. The scope of study was mainly to evaluate the neutron irradiation induced change in nil ductility transition temperature and tensile properties on some selected materials. The required quantity of these materials was received from Japan, France, Federal Republic of Germany and United States of America. The paper reports the results obtained from the investigations carried out in the Radiometallurgy Division of Bhabha Atomic Research Centre, Trombay, Bombay, India, on Japanese plate, Japanese forging, French plate, French forging and French weld. Due to other important commitments of the CIRUS reactor, in which the irradiation was carried out, it was not possible to complete the entire programme at this time. (author)

  8. The disposal of reactor pressure vessels heads at the Soulaines LLW disposal site (France)

    International Nuclear Information System (INIS)

    Lecoq, Pascal; Dutzer, Michel

    2012-01-01

    In 1994, EDF, the national utility, requested ANDRA (Agence nationale pour la gestion des dechets radioactifs), the French national agency for the management of radioactive waste, to take over 55 reactor pressure vessel (RPV) heads from all PWRs (900 and 1,300 MW). By 1994, 24 of the 55 vessel heads had already been removed and stored in an EDF facility located in southern France. Since the disposal of such large components was not described and licensed in the safety report of the Centre de l'Aube and in the regulatory body's technical prescriptions, it was necessary to request a special authorisation from the Nuclear Safety Authority (Autorite de surete nucleaire - ASN), which is the French regulatory body. The study relating to the disposal of large components (long-term safety and operation aspects) began in 1995, and the first authorisation request was sent to ASN in December 1997. The main provisions being foreseen included: the disposal of the RPV head with the use of a biological shielding plate under the head and of a protection plate for adapters; the disposal in dedicated vaults; the acceptance of RPV heads disposal according to their radiological characterisation. Investigations by the regulatory body lasted 18 months, and the decision was unfavourable, due to the lack of an envelope (packaging) all around the waste, and the observation, with regard to the embedding conditions, of a continuity between the package and the filling material (inside the vault), without that continuous envelope. Consequently, the file was revised: the package concept was changed, and the confining envelope, which was initially used for transportation only, was maintained for disposal purposes in order to form, together with the shielding plate, the first barrier around the waste. Hence, the mass of the large components increased from 60 and 80 t to 100 and 120 t, respectively. The delivery of vessel heads at the Soulaines disposal site was discussed during a meeting of the

  9. SURROGATE MODEL DEVELOPMENT AND VALIDATION FOR RELIABILITY ANALYSIS OF REACTOR PRESSURE VESSELS

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, William M.; Riley, Matthew E.; Spencer, Benjamin W.

    2016-07-01

    In nuclear light water reactors (LWRs), the reactor coolant, core and shroud are contained within a massive, thick walled steel vessel known as a reactor pressure vessel (RPV). Given the tremendous size of these structures, RPVs typically contain a large population of pre-existing flaws introduced in the manufacturing process. After many years of operation, irradiation-induced embrittlement makes these vessels increasingly susceptible to fracture initiation at the locations of the pre-existing flaws. Because of the uncertainty in the loading conditions, flaw characteristics and material properties, probabilistic methods are widely accepted and used in assessing RPV integrity. The Fracture Analysis of Vessels – Oak Ridge (FAVOR) computer program developed by researchers at Oak Ridge National Laboratory is widely used for this purpose. This program can be used in order to perform deterministic and probabilistic risk-informed analyses of the structural integrity of an RPV subjected to a range of thermal-hydraulic events. FAVOR uses a one-dimensional representation of the global response of the RPV, which is appropriate for the beltline region, which experiences the most embrittlement, and employs an influence coefficient technique to rapidly compute stress intensity factors for axis-aligned surface-breaking flaws. The Grizzly code is currently under development at Idaho National Laboratory (INL) to be used as a general multiphysics simulation tool to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled RPVs. Grizzly can be used to model the thermo-mechanical response of an RPV under transient conditions observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local 3D models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in

  10. Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

  11. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1987-01-01

    This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 10 8 per m 3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

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

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

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

  13. Fabrication Flaw Density and Distribution in the Repairs of Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    Schuster, George J.; Doctor, Steven R.; Simonen, Fredric A.

    2006-01-01

    The Pacific Northwest National Laboratory (PNNL) is developing a generalized flaw size and density distribution for the population of U.S. reactor pressure vessels (RPVs). The purpose of the generalized flaw distribution is to predict vessel specific flaw rates for use in probabilistic fracture mechanics calculations that estimate vessel failure probability. Considerable progress has been made on the construction of an engineering data base of fabrication flaws in U.S. nuclear RPVs. The fabrication processes and product forms used to construct U.S. RPVs are represented in the data base. A validation methodology has been developed for characterizing the flaws for size, shape, orientation, and composition. The relevance of construction records has been established for describing fabrication processes and product forms. The fabrication flaws were detected in material removed from cancelled nuclear power plants using high sensitivity nondestructive ultrasonic testing, and validated by other nondestructive evaluation (NDE) techniques, and complemented by destructive testing. This paper describes research that has generated data on welding flaws, which indicated that the largest flaws occur in weld repairs. Recent research results confirm that repair flaws are complex in composition and may include cracks on the repair ends. Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for nuclear power plant components requires radiographic examinations (RT) of welds and requires repairs for RT indications that exceed code acceptable sizes. PNNL has previously obtained the complete construction records for two RPVs. Analysis of these records show a significant change in repair frequency.

  14. Experimental evaluation of surveillance capsule assemblies for life assessment of CHASNUPP Unit-1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Şahin, Sümer; Saeed, Asim

    2016-01-01

    Highlights: • Disassembling of dose and flux boxes in glove box. • Activity measurement of radiometric sensors using HPGE γ-spectrometer. • Decay corrected reaction rate calculation using measured activity data and plant power history. • Best estimation of fast neutron flux using LSL-M2 spectrum unfolding code. • Comparison with transport calculations performed using DOT3.5. - Abstract: Neutron flux and energy spectrum were determined at the surface of three in-vessel Surveillance Capsule Assemblies (SCAs) removed from CHASNUPP Unit-1 after 2nd, 4th, and 9th fuel cycles for the life assessment of reactor pressure vessel belt line region. Dosimetry data were measured from radiometric sensors irradiated in base material section of SCAs. Fast neutron flux (E > 1.0 MeV) was best estimated at the surface of three SCAs corresponding to the center of C-1 core using the least square method by employing LSL-M2 package. These results were compared with fast neutron flux calculated using DOT3.5 code and both results are within good agreement of ±20% acceptance criteria as described in Regulatory Guide 1.190. Therefore, calculational model was validated by dosimetry evaluation and these results can be used in the life assessment of CHASNUPP Unit-1 pressure vessel belt line region.

  15. Solid-state track recorder neutron dosimetry in light water reactor pressure vessel surveillance mockups

    International Nuclear Information System (INIS)

    Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

    1984-09-01

    Solid-State Track Recorder (SSTR) measurements of neutron-induced fission rates have been made in several pressure vessel mockup facilities as part of the US Nuclear Regulatory Commission's (NRC) Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP). The results of extensive physics-dosimetry measurements made at the Pool Critical Assembly (PCA) at Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN are summarized. Included are 235 U, 238 U, 237 Np and 232 Th fission rates in the PCA 12/13, 8/7, and 4/12 SSC configurations. Additional low power measurements have been made in an engineering mockup at the VENUS critical assembly at CEN-SCK, Mol, Belgium. 237 Np and 238 U fission rates were made at selected locations in the VENUS mockup, which models the in-core and near-core regions of a pressurized water reactor (PWR). Absolute core power measurements were made at VENUS by exposing solid-state track recorders (SSTRs) to polished fuel pellets within in-core fuel pins. 8 references, 4 figures, 10 tables

  16. Stress and strain analysis in the plastic fields of a PWR pressure vessel

    International Nuclear Information System (INIS)

    Chevalier, G.; Clement, G.; Goldstein, S.; Hoffman, A.

    1975-01-01

    The pressure vessel of the pressurized water reactor project Champlain has been calculated twice with two different programs of the CEA-SEMT system: the first axisymmetrical and the second on shell elements. The calculations involved the same load diagram and material of the same characteristics represented by the uniaxial traction test on a sample of the steel used for this vessel. An axisymmetrical calculation with the PASTEL program was carried out on a section of the pressure vessel between two inlet points, which means that the structure considered is assumed as cylindrical. A shell element calculation with the TRICO program was carried out on a quarter vessel containing an inlet point of 700 mm diameter. The results obtained with these two CEA programs agree very well. It was possible to show how the plasticity progresses with each pressure step. These two studies enabled the excess overall deformation to be estimated under a load 2.5 times the nominal pressure, this deformation appearing mainly in the cylindrical part of the vessel. It was shown that at such loads no plastic instability occurs. (Auth.)

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

    International Nuclear Information System (INIS)

    Francisco, Alexandre S.; Duran, Jorge Alberto R.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Habil; Ziliukas, A.

    1997-01-01

    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

  19. Alternative welding reconditioning solutions without post welding heat treatment of pressure vessel

    Science.gov (United States)

    Cicic, D. T.; Rontescu, C.; Bogatu, A. M.; Dijmărescu, M. C.

    2017-08-01

    In pressure vessels, working on high temperature and high pressure may appear some defects, cracks for example, which may lead to failure in operation. When these nonconformities are identified, after certain examination, testing and result interpretation, the decision taken is to repair or to replace the deteriorate component. In the current legislation it’s stipulated that any repair, alteration or modification to an item of pressurised equipment that was originally post-weld heat treated after welding (PWHT) should be post-weld heat treated again after repair, requirement that cannot always be respected. For that reason, worldwide, there were developed various welding repair techniques without PWHT, among we find the Half Bead Technique (HBT) and Controlled Deposition Technique (CDT). The paper presents the experimental results obtained by applying the welding reconditioning techniques HBT and CDT in order to restore as quickly as possible the pressure vessels made of 13CrMo4-5. The effects of these techniques upon the heat affected zone are analysed, the graphics of the hardness variation are drawn and the resulted structures are compared in the two cases.

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

    Energy Technology Data Exchange (ETDEWEB)

    Habil; Ziliukas, A.

    1997-04-01

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

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

    International Nuclear Information System (INIS)

    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 (-60 degree C). 21 refs., 5 figs., 3 tabs

  2. Fabrication of High Temperature and High Pressure Vessel for the Fuel Test

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Sim, Bong Shick; Shon, Jae Min; Ahn, Seung Ho; Yoo, Seong Yeon

    2007-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR and CANDU nuclear power plants has been developed and installed in HANARO, KAERI. It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is located inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

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

    2017-01-01

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

  5. Development of automated ultrasonic device for in-service inspection of ABWR pressure vessel bottom head

    International Nuclear Information System (INIS)

    Kojima, Y.; Matsuyama, A.

    1995-01-01

    An automated device and its controller have been developed for the bottom head weld examination of pressure vessel of Advanced Boiling Water Reactor (ABWR). The internal pump casings and the housings of control rod prevent a conventional ultrasonic device from scanning the required inspection zone. With this reason, it is required to develop a new device to examine the bottom head area of ABWR. The developed device is characterized by the following features. (1) Composed of a mother vehicle and a compact inspection vehicle. They are connected only by an electric wire without using the conventional arm mechanism. (2) The mother vehicle travels on a track and lift up the inspection vehicle to the vessel. (3) The mother vehicle can automatically attach the inspection vehicle to the bottom head, and detach the inspection vehicle from it. (4) Collision avoidance control function with a touch sensor is installed at the front of the inspection vehicle. The device was successfully demonstrated using a mock-up of reactor pressure vessel

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

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

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

    Science.gov (United States)

    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.

  8. Ultrasonic phased array examination of circumferential weld joint in reactor pressure vessel of BWR

    International Nuclear Information System (INIS)

    Nanekar, Paritosh; Jothilakshmi, N.; Jayakumar, T.

    2013-01-01

    Highlights: • Phased array technique developed for weld joint inspection in BWR pressure vessel. • Simulation studies were carried out for conventional and phased array probe. • Conventional ultrasonic test shows in-adequate weld coverage and poor resolution. • Focused sound beam in phased array results in good resolution and sensitivity. • Ultrasonic phased array technique is validated on mock-up with reference defects. - Abstract: The weld joints in the reactor pressure vessel (RPV) of Boiling Water Reactors (BWR) are required to be examined periodically for assurance of structural integrity. Ultrasonic phased array examination technique has been developed in authors’ laboratory for inspection of the top flange to shell circumferential weld joint in RPV of BWRs, which are in operation in India since the late 1960s. The development involved detailed simulation studies for computation of focal laws followed by validation on mock-up. The paper brings out the limitations of the conventional ultrasonic technique and how this can be overcome by the phased array approach for the weld joint under consideration. The phased array technique was successfully employed for field examination of this weld joint in RPV during the re-fuelling outage

  9. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

  10. Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

    International Nuclear Information System (INIS)

    Krasikov, E.A.; Nikolaenko, V.A.

    2008-01-01

    Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

  11. Models for ductile crack initiation and tearing resistance under mode 1 loading in pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, M.R.

    1988-06-01

    Micromechanistic models are presented which aim to predict plane strain ductile initiation toughness, tearing resistance and notched bar fracture strains in pressure vessel steels under monotonically increasing tensile (mode 1) loading. The models for initiation toughness and tearing resistance recognize that ductile fracture proceeds by the growth and linkage of voids with the crack-tip. The models are shown to predict the trend of initiation toughness with inclusion spacing/size ratio and can bound the available experimental data. The model for crack growth can reproduce the tearing resistance of a pressure vessel steel up to and just beyond crack growth initiation. The fracture strains of notched bars pulled in tension are shown to correspond to the achievement of a critical volume fraction of voids. This criterion is combined with the true stress - true strain history of a material point ahead of a blunting crack-tip to predict the initiation toughness. An attempt was made to predict the fracture strains of notched tensile bars by adopting a model which predicts the onset of a shear localization phenomenon. Fracture strains of the correct order are computed only if a ''secondary'' void nucleation event at carbide precipitates is taken into account. (author)

  12. Evaluation of the role of Cr, Ni, Mn and Si in reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Degmova, J.; Veternikova, J.; Sojak, S.; Slugen, V. [Department of Nuclear Physics and Technology, Faculty of Electrical Engineering and Information Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia); Krsjak, V.; Acosta, B. [JRC-IE, Joint Research Centre, Institute for Energy, P.O. Box 2, Petten 1755 ZG (Netherlands)

    2010-07-01

    In this paper we report initial results from a study focused on investigation of the role of elements as Cr, Ni, Mn and Si in the radiation stability of reactor pressure vessel steels. Twelve model ferritic steels with basic composition derived from Russian WWER-1000 and Western PWR reactor pressure vessel materials were studied by Charpy-V impact, magnetic Barkhausen noise, Vickers hardness tests, Relative Seebeck Coefficient measurements and Positron annihilation spectroscopy. Higher Cr content in model steels was found generally to give increased root mean square values independent of Mn and Si contents. The ductile-brittle transition temperatures and hardness values of the model steels were found to be independent of composition. The correlation between ductile- brittle transition temperatures and hardness values has potential for prompt determination of the effect of composition and irradiation on the steel properties. The next stage of the assessment will study the effect of irradiation of the model steels to accumulated neutron fluences of 10{sup 19} n.cm{sup -2}. (authors)

  13. Summary of the NEA/CSNI Workshop on the safety assessment of reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Toerroenen, K. (VTT (Finland)); English, C.A. (AEA Tehnology (United Kingdom)); Crutzen, S. (CEC/TRC (Italy)); Hemsworth, B. (Health and Safety Executive (United Kingdom)); Noel, R.C. (EDF (France))

    1990-12-01

    The workshop was initiated by the CSNI Principal Working Group No 3, Reactor Component Integrity as part of its programme concerned with the long term degradation and service life management of reactor components. A central issue in this context and in consideration of reactor plant license extension is the capability to provide reliable safety assessments of component integrity using available data and assessment methods. In this regard, The Principal Working Group plans a systematic survey of reactor structural components critical to safety, commencing with pressure vessels. Existing knowledge and recent achievements in research on the behaviour of neutron irradiated materials, applications of fracture mechanics and probabilistic safety assessment methods provide comprehensive information to the regulatory bodies and plant operators. The Group considered that a workshop penetrating the problems met by authorities, utilities and researchers assessing the safety of reactor pressure vessels would clarify the current state of knowledge and the need and direction for further research. This report is a summary of the NEA/CSNI workshop. (au).

  14. New developments in Nickel-Hydrogen Dependent Pressure Vessel (DPV) cell and battery design

    Science.gov (United States)

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

    1997-01-01

    The Dependent Pressure Vessel (DPV) Nickel-Hydrogen (NiH2) design is being developed by Eagle-Picher Industries, Inc. (EPI) as an advanced battery for military and commercial, aerospace and terrestrial applications. The DPV cell design offers high specific energy, energy density and reduced cost, while retaining the established Individual Pressure Vessel (IPV) technology flight heritage and database. This advanced DPV design also offers a more efficient mechanical, electrical and thermal cell and battery configuration and a reduced parts count. The DPV battery design promotes compact, minimum volume packaging and weight efficiency, and delivers significant cost and weight savings while providing minimal design risks. This presentation will discuss new DPV design concepts and production features and present test data from existing development cells. The DPV combines the unique features and significant advantages of NiH2 electrochemistry with the simplicity and extensive design heritage of the NiCd battery system to create a power source ideal for military and commercial terrestrial projects as well as traditional aerospace applications.

  15. Acoustoelastic evaluation of welding and heat treatment stress relieving of pressure vessel steel for Angra 3

    Energy Technology Data Exchange (ETDEWEB)

    Moraes, Bruno C. de, E-mail: bruno.cesar@nuclep.gov.br [Nuclebras Equipamentos Pesados S.A (NUCLEP), Itaguai, RJ (Brazil); Bittencourt, Marcelo de S.Q., E-mail: bruno.cesar@nuclep.gov.br, E-mail: bittenc@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Currently the knowledge of non-destructive techniques allows to evaluate the stresses on components and mechanical structures, aiming at physical security, preservation of the environment and avoid financial losses associated with the construction and operation of industrial plants. The search for new techniques, especially applied in the nuclear industry to assess status more accurately, voltage safety and to ensure structural integrity, for example, core components of the primary circuit, such as the reactor pressure vessel and steam generator has become of great importance within the community of non-destructive testing .This paper aims to contribute to the non-destructive technique development in order to ensure the structural integrity of nuclear components. One acoustoelastic evaluation of steel 20 MnMoNi 55, used in pressure vessels of nuclear power plants were performed. The acoustic birefringence technique was use to evaluate the acoustoelastic behavior of the test material in the as received condition, after welding and after the stress relief heat treatment. The constant acoustoelastic material was obtained by an uniaxial loading test. It was found a slight anisotropy in the material as received. After welding, a marked variation of acoustic birefringence in the region near the weld bead was observed. The heat treatment indicated a new change of acoustic birefringence. Obtaining the acoustoelastic constant allowed the evaluation of stress in the different conditions of the weld and treated material. (author)

  16. A mathematical model for pressure-based organs behaving as biological pressure vessels.

    Science.gov (United States)

    Casha, Aaron R; Camilleri, Liberato; Gauci, Marilyn; Gatt, Ruben; Sladden, David; Chetcuti, Stanley; Grima, Joseph N

    2018-04-26

    We introduce a mathematical model that describes the allometry of physical characteristics of hollow organs behaving as pressure vessels based on the physics of ideal pressure vessels. The model was validated by studying parameters such as body and organ mass, systolic and diastolic pressures, internal and external dimensions, pressurization energy and organ energy output measurements of pressure-based organs in a wide range of mammals and birds. Seven rules were derived that govern amongst others, lack of size efficiency on scaling to larger organ sizes, matching organ size in the same species, equal relative efficiency in pressurization energy across species and direct size matching between organ mass and mass of contents. The lung, heart and bladder follow these predicted theoretical relationships with a similar relative efficiency across various mammalian and avian species; an exception is cardiac output in mammals with a mass exceeding 10kg. This may limit massive body size in mammals, breaking Cope's rule that populations evolve to increase in body size over time. Such a limit was not found in large flightless birds exceeding 100kg, leading to speculation about unlimited dinosaur size should dinosaurs carry avian-like cardiac characteristics. Copyright © 2018. Published by Elsevier Ltd.

  17. Design of Semi-composite Pressure Vessel using Fuzzy and FEM

    Science.gov (United States)

    Sabour, Mohammad H.; Foghani, Mohammad F.

    2010-04-01

    The present study attempts to present a new method to design a semi-composite pressure vessel (known as hoop-wrapped composite cylinder) using fuzzy decision making and finite element method. A metal-composite vessel was designed based on ISO criteria and then the weight of the vessel was optimized for various fibers of carbon, glass and Kevlar in the cylindrical vessel. Failure criteria of von-Mises and Hoffman were respectively employed for the steel liner and the composite reinforcement to characterize the yielding/ buckling of the cylindrical pressure vessel. The fuzzy decision maker was used to estimate the thickness of the steel liner and the number of composite layers. The ratio of stresses on the composite fibers and the working pressure as well as the ratio of stresses on the composite fibers and the burst (failure) pressure were assessed. ANSYS nonlinear finite element solver was used to analyze the residual stress in the steel liner induced due to an auto-frettage process. Result of analysis verified that carbon fibers are the most suitable reinforcement to increase strength of cylinder while the weight stayed appreciably low.

  18. Performance of asphaltic concrete incorporating styrene butadiene rubber subjected to varying aging condition

    Science.gov (United States)

    Salah, Faisal Mohammed; Jaya, Ramadhansyah Putra; Mohamed, Azman; Hassan, Norhidayah Abdul; Rosni, Nurul Najihah Mad; Mohamed, Abdullahi Ali; Agussabti

    2017-12-01

    The influence of styrene butadiene rubber (SBR) on asphaltic concrete properties at different aging conditions was presented in this study. These aging conditions were named as un-aged, short-term, and long-term aging. The conventional asphalt binder of penetration grade 60/70 was used in this work. Four different levels of SBR addition were employed (i.e., 0 %, 1 %, 3 %, and 5 % by binder weight). Asphalt concrete mixes were prepared at selected optimum asphalt content (5 %). The performance was evaluated based on Marshall Stability, resilient modulus, and dynamic creep tests. Results indicated the improving stability and permanent deformation characteristics that the mixes modified with SBR polymer have under aging conditions. The result also showed that the stability, resilient modulus, and dynamic creep tests have the highest rates compared to the short-term aging and un-aged samples. Thus, the use of 5 % SBR can produce more durable asphalt concrete mixtures with better serviceability.

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

    International Nuclear Information System (INIS)

    Grounes, M.

    1966-03-01

    After a short review of the part of the Swedish nuclear energy program that is of interest in this context the Swedish reactor pressure vessels and the reasoning behind the choice of materials are surveyed. Problems and desirable aims for future reactors are discussed. Much work is now being done on new types of pressure vessel steels with high strength, low transition temperature and good corrosion resistance. These steels are of the martensitic austenitic type Bofors 2RMO (13 % Cr, 6 % Ni, 1. 5 % Mo) and of the ferritic martensitic austenitic type Avesta 248 SV (16 % Cr, 5 % Ni, 1 % Mo). An applied philosophy for estimating the brittle-fracture tendency of pressure vessels is described. As a criterion of this tendency we use the crack-propagation transition temperature, e. g. as measured by the Robertson isothermal crack-arrest test. An estimate of this transition temperature at the end of the reactor' s lifetime must take increases due to fabrication, welding, geometry, ageing and irradiation into account. The transition temperature vs. stress curve moves towards higher temperatures during the reactor' s lifetime. As long as this curve does not cross the reactor vessel stress vs. temperature curve the vessel is considered safe. The magnitude of the different factors influencing the final transition temperature are discussed and data for the Marviken reactor's pressure vessel are presented. At the end of the reactor's lifetime the estimated transition temperature is 115 deg C, which is below the maximum permissible value. A program for the study of strain ageing has been initiated owing to the uncertainty as to the extent of strain ageing at low strains. A study of a simple crack-arrest test, developed in Sweden, is in progress. An extensive irradiation-effects program on several steels is in progress. Results from tests on the Swedish carbon-manganese steels 2103/R3, SIS 142103 and SIS 142102, the low-alloy steels Degerfors DE-631A, Bofors NO 345 and Fortiweld

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

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M.

    1966-03-15

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