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Sample records for concrete pressure vessel

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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.)

  6. 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)

  7. 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)

  8. 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

  9. 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

  10. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. 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)

  12. 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.)

  13. 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.)

  14. 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

  15. 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

  16. 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

  17. 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.)

  18. 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

  19. 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

  20. 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

  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. 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.

  4. 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

  5. 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

  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. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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.)

  13. 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)

  14. 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)

  15. 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.)

  16. 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

  17. 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

  18. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    Prediction of the response of the Sandia National laboratory 1/6-scale reinforced concrete containment model test was obtained by Argonne National Laboratory (ANL) employing a computer program developed by ANL. The test model was internally pressurized to failure. The two-dimensional code TEMP-STRESS [1-5] has been developed at ANL for stress analysis of plane and axisymmetric 2-D reinforced structures under various thermal conditions. The program is applicable to a wide variety of nonlinear problems, and is utilized in the present study. The comparison of these pretest computations with test data on the containment model should be a good indication of the state of the code

  19. 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

  20. 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

  1. 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)

  2. Stress criteria for nuclear vessel concrete

    International Nuclear Information System (INIS)

    Costes, D.

    1975-01-01

    Concrete nuclear vessels are submitted to prestressing forces which limit tensile stresses in concrete when the vessel is under pressure with thermal gradients. Hence, the most severe conditions for concrete appear when the vessel is prestressed and not submitted to internal pressure. The triaxial states of stress in the concrete may be computed postulating elastic or other behavior and compared with safe limits obtained from rupture tests and fatigue tests. The first part of the paper, recalls experimental rupture results and the acceptability procedures currently used. Criteria founded on the lemniscoid surfaces are proposed, parameters for which are obtained by various tests and safety considerations. In the second part, rupture tests are reported on small, thick, cylindrical vessels submitted to external hydraulic pressure simulating prestressing forces. Materials used are plain concrete, microconcrete, marble and graphite. The strengths obtained are much higher than those which could be elastically computed, triaxial rupture states being provided by previous experiments. Such results may be due to a plastic stress redistribution before fracture and to stabilizing effects of stress gradients around the more stressed areas. Fatigue tests by external hydraulic loading are reported [fr

  3. Analysis study on change of tendon behavior during pressurization process of Pre-stressed Concrete Containment Vessel

    International Nuclear Information System (INIS)

    Kashiwase, Takako; Nagasaka, Hideo

    1999-01-01

    NUPEC has been planning the ultimate strength test of Pre-stressed Concrete Containment Vessel (PCCV). The test model is 1/4 uniform scale model of Japan actual PCCV. It involves an equipment hatch, several penetrations and liner with T-anchors. The ancillary test for the PCCV test was conducted, in which friction coefficient of hoop tendon was evaluated by tensile force distribution using the same tendon as that of 1/4 PCCV model. Tendon will be in plastic region under internal pressure above 3.5 times design pressure (Pd) and surface characteristic of tendon and the resultant friction coefficient will be changed. In the present paper, tendon friction coefficient in the plastic region was obtained by evaluating plastic region data of tendon in the ancillary test. The validity of the obtained friction coefficient was confirmed by the tendon elongation data. In addition to the formally developed elastic region friction coefficient, the obtained plastic region correlation was incorporated into ABAQUS Ver. 5.6. The effect of tendon tensile force distribution change on structural behavior up to 3.8 Pd was evaluated. (author)

  4. 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 ...

  5. 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

  6. 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

  7. 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)

  8. 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.)

  9. PRESSURE-RESISTANT VESSEL

    NARCIS (Netherlands)

    Beukers, A.; De Jong, T.

    1997-01-01

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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

    International Nuclear Information System (INIS)

    Oland, C.B.

    1985-01-01

    Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

  15. 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

  16. 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.

  17. 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

  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. 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.

  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. 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)

  3. 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.

  4. High pressure storage vessel

    Science.gov (United States)

    Liu, Qiang

    2013-08-27

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

  5. Instrumentation and testing of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Pace, D.W.; Klamerus, E.W.

    1997-01-01

    Static overpressurization tests of two scale models of nuclear containment structures - a steel containment vessel (SCV) representative of an improved, boiling water reactor (BWR) Mark II design and a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR) - are being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. This paper discusses plans for instrumentation and testing of the PCCV model. 6 refs., 2 figs., 2 tabs

  6. A thermal insulation system intended for a prestressed concrete vessel

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1975-01-01

    The description is given of a thermal insulation system withstanding the pressure of a vaporisable fluid for a prestressed concrete vessel, particularly the vessel of a boiling water nuclear reactor. The ring in the lower part of the vessel has, between the fluid inlet pipes and the bottom of the vessel, an annular opening of which the bottom edge is integral with an annular part rising inside the ring and parallel to it. This ring is hermetically connected to the bottom of the vessel and is coated with a metal lagging, at least facing the annular opening. This annular opening is made in the ring half-way up between the fluid inlet pipes and the bottom of the vessel. It is connected to the bottom of the vessel through the internal structure enveloping the reactor core [fr

  7. Review of concrete properties for prestressed concrete pressure vesssels

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, R.K.

    1976-10-01

    The desire for increasing power output along with safety requirements has resulted in consideration of the prestressed concrete pressure vessel (PCPV) for most current nuclear reactor systems, as well as for the very-high-temperature reactor for process heat and as primary pressure vessels for coal conversion systems. Results are presented of a literature review to ascertain current knowledge regarding plain concrete properties under conditions imposed by a mass concrete structure such as PCRV. The effects of high temperature on such properties as strength, elasticity, and creep are discussed, as well as changes in thermal properties, multiaxial behavior, and the mechanisms thought to be responsible for the observed behavior. In addition, the effects of radiation and moisture migration are discussed. It is concluded that testing results found in the technical literature show much disagreement as to the effects of temperature on concrete properties. The variations in concrete mixtures, curing and testing procedures, age at loading, and moisture conditions during exposure and testing are some of the reasons for such disagreement. Test results must be limited, in most cases, to the materials and conditions of a given test rather than applied to such a general class of materials such as concrete. It is also concluded that sustained exposure of normal concretes to current PCRV operating conditions will not result in any significant loss of properties. However, lack of knowledge regarding effects of temperatures exceeding 100/sup 0/C (212/sup 0/F), moisture migration, and multiaxial behavior precludes a statement advocating operation beyond current design limits. The report includes recommendations for future research on concrete for PCPVs.

  8. 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

  9. Requirements for thermal insulation on prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Neylan, A.J.; Wistrom, J.D.

    1979-01-01

    During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

  10. Nonlinear analysis of end slabs in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.

    1978-01-01

    A procedure for the nonlinear analysis of end slabs is prestressed concrete reactor vessels (PCRVs), based on the finite element method, is presented. The applicability of the procedure to the ultimate load analysis of small-scale models of the primary containment of nuclear reactors is shown. Material nonlinearity only is considered. The procedure utilizes the four-node linear quadrilateral isoparametric element with the choice of incorporating the nonconforming modes. This element is used for modeling the vessel as an axisymmetric solid. Concrete is assumed to be an isotropic material in the elastic range. The compressive stresses are judged according to a special form of the Mohr-Coulomb criterion. The nonlinear problem was solved using a generalized Newton-Raphson procedure. A detailed example problem of a pressure vessel with penetrations is presented. This is followed by a summary of the other cases studied. The solutions obtained match very closely the measured response of the test vessels under increasing internal pressure up to failure. The procedure is thus adequate for the assessment of the ultimate load behavior and failure of actual pressure vessels with a moderate demand on human and computational resources

  11. Development of prestressed concrete containment vessels

    International Nuclear Information System (INIS)

    Yuji, Hideo; Kuniyoshi, Mutsumu; Nagata, Kaoru

    1983-01-01

    This paper presents a summary of evaluations for the selection of the structural and prestressing system type to be employed for the first domestic Prestressed Concrete Containment Vessel (PCCV) in Japan. This paper also discusses characteristic features in the design of the liner plate system provided on the PCCV inner surface to assure its leak-tight integrity. Prestressed concrete containment vessels so far constructed in foreign countries are to a considerable extent of different structural types, depending on differences in dome shapes, prestressing systems and number of buttresses. These differences are caused not only by differences in design philosophy and construction practices, but also by difference in the level of technology of the times when the individual containment vessels are being constructed. In the investigation reported herein, the most suitable types of PCCV and Prestressing Systems were determined as the results of an overall comparative evaluation of data and information obtained from PCCV's so far constructed from the design, construction and cost aspects, taking into consideration the seismic criteria, available technology, construction practices, regulations and technical standards in Japan. The function of the liner plate system requires the liner to have enough deformability so that the liner deformation can be consistent with the PCCV concrete deformation. Therefore, in the design of the liner plate system a method for evaluating liner deformability was employed, instead of the stress evaluation method which is widely used in the design of ordinary structures. (author)

  12. Material problems in accident analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1977-01-01

    Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

  13. Study on prestressed concrete reactor vessel structures. II-5: Crack analysis by three dimensional finite elements method of 1/20 multicavity type PCRV subjected to internal pressure

    Science.gov (United States)

    1978-01-01

    A three-dimensional finite elements analysis is reported of the nonlinear behavior of PCRV subjected to internal pressure by comparing calculated results with test results. As the first stage, an analysis considering the nonlinearity of cracking in concrete was attempted. As a result, it is found possible to make an analysis up to three times the design pressure (50 kg/sqcm), and calculated results agree well with test results.

  14. Construction of reactor vessel bottom of prestressed reinforced concrete

    International Nuclear Information System (INIS)

    Sitnikov, M.I.; Metel'skij, V.P.

    1980-01-01

    Methods are described for building reactor vessel bottoms of prestressed reinforced concrete during NPPs construction in Great Britain, France, Germany (F.R.) and the USA. Schematic of operations performed in succession is presented. Considered are different versions of one of the methods for concreting a space under a facing by forcing concrete through a hole in the facing. The method provides tight sticking of the facing to the reactor vessel bottom concrete

  15. 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

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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...

  3. 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...

  4. 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...

  5. 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)

  6. Crack analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Gallix, R.; Liu, T.C.; Lu, S.C.H.

    1975-01-01

    A new method to perform the crack analysis of non-axisymmetric, multicavity prestressed concrete reactor vessels (PCRV's) subjected to hypothetical overpressure by using an axisymmetric two-dimensional finite element computer code is presented. Concrete, steel liner, bonded reinforcing steel and prestressing steel elements are modeled. The limiting tensile strain criterion is adopted for concrete cracking. The steel elements are assumed to be elastic/perfectly plastic. Von Mises yield criterion and Prandtl-Reuss flow equations define the behavior of the liner in the range of plastic deformations. An orthotropic stress-strain constitutive law is utilized for cracked concrete elements. To account for the presence of penetrations and secondary cavities in the PCRV, a modified finite element model based on the concept of effective moduli is adopted. The pressure in these cavities is simulated by equivalent axisymmetric pressure distributions. In the analysis, the pressure is applied incrementally. For a given pressure, the displacements, strains, and stresses are computed. The state of strains or stresses is then examined against the cracking or yield criteria. If cracking or yield is indicated, the stiffness and load matrices for the cracked and yielding elements are recomputed and a new equilibrium is sought. This procedure is repeated until the desired convergence of the solution is achieved. The validity of the adopted approach utilizing the two-dimensional finite element method for overpressure analyses of non-axisymmetric PCRV's is demonstrated through comparisons with two multicavity PCRV scale models. A reliable and conservative estimate of PCRV behavior under overpressure is obtained

  7. 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.

  8. 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.

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

    Science.gov (United States)

    2010-10-01

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

  10. 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

  11. Seismic proof test of a reinforced concrete containment vessel (RCCV)

    International Nuclear Information System (INIS)

    Hirama, Toshihiko; Goto, Masashi; Shiba, Keiji; Kobayashi, Toshio; Tanaka, Ryozo; Tsurumaki, Shizuo; Takiguchi, Katsuki; Akiyama, Hiroshi

    2005-01-01

    A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant. The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load-deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4

  12. 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.

  13. Instrumentation of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Rightley, M.J.; Matsumoto, T.

    1995-01-01

    A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission. At present, two tests are being planned: a test of a model of a steel containment vessel (SCV) that is representative of an improved, boiling water reactor (BWR) Mark II design; and a test of a model of a prestressed concrete containment vessel (PCCV). This paper discusses plans and the results of a preliminary investigation of the instrumentation of the PCCV model. The instrumentation suite for this model will consist of approximately 2000 channels of data to record displacements, strains in the reinforcing steel, prestressing tendons, concrete, steel liner and liner anchors, as well as pressure and temperature. The instrumentation is being designed to monitor the response of the model during prestressing operations, during Structural Integrity and Integrated Leak Rate testing, and during test to failure of the model. Particular emphasis has been placed on instrumentation of the prestressing system in order to understand the behavior of the prestressing strands at design and beyond design pressure levels. Current plans are to place load cells at both ends of one third of the tendons in addition to placing strain measurement devices along the length of selected tendons. Strain measurements will be made using conventional bonded foil resistance gages and a wire resistance gage, known as a open-quotes Tensmegclose quotes reg-sign gage, specifically designed for use with seven-wire strand. The results of preliminary tests of both types of gages, in the laboratory and in a simulated model configuration, are reported and plans for instrumentation of the model are discussed

  14. Design and analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

    1977-01-01

    During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels

  15. Pretest Round Robin Analysis of 1:4-Scale Prestressed Concrete Containment Vessel Model

    Energy Technology Data Exchange (ETDEWEB)

    HESSHEIMER,MICHAEL F.; LUK,VINCENT K.; KLAMERUS,ERIC W.; SHIBATA,S.; MITSUGI,S.; COSTELLO,J.F.

    2000-12-18

    The purpose of the program is to investigate the response of representative scale models of nuclear containment to pressure loading beyond the design basis accident and to compare analytical predictions to measured behavior. This objective is accomplished by conducting static, pneumatic overpressurization tests of scale models at ambient temperature. This research program consists of testing two scale models: a steel containment vessel (SCV) model (tested in 1996) and a prestressed concrete containment vessel (PCCV) model, which is the subject of this paper.

  16. Analytical model for shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.; Sozen, M.A.; Schnobrich, W.C.

    1979-04-01

    The results are presented of an investigation of the behavior and strength of flat end slabs of cylindrical prestressed concrete nuclear reactor vessels. The investigation included tests of ten small-scale pressure vessels and development of a nonlinear finite-element model to simulate the deformation response and strength of the end slabs. Because earlier experimental studies had shown that the flexural strength of the end slab could be calculated using intelligible procedures, the emphasis of this investigation was on shear strength

  17. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  18. Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Young Sun; Park, Hyung Kui [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcement was shown to be more effective at a high pressure loading and a low prestress level.

  19. Containment performance evaluation of prestressed concrete containment vessels with fiber reinforcement

    International Nuclear Information System (INIS)

    Choun, Young Sun; Park, Hyung Kui

    2015-01-01

    Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcement was shown to be more effective at a high pressure loading and a low prestress level

  20. Prestressed concrete reactor vessel thermal cylinder model study

    International Nuclear Information System (INIS)

    Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.

    1977-01-01

    The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a 1 / 6 -scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating

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

    Science.gov (United States)

    Yi, Ping; Wang, Qingkang; Kong, Xianjing

    2017-01-01

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

  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. 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)

  4. 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

  5. Lightweight bladder lined pressure vessels

    Science.gov (United States)

    Mitlitsky, Fred; Myers, Blake; Magnotta, Frank

    1998-01-01

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

  6. 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.)

  7. Emergency venting of pressure vessels

    International Nuclear Information System (INIS)

    Steinkamp, H.

    1995-01-01

    With the numerical codes developed for safety analysis the venting of steam vessel can be simulated. ATHLET especially is able to predict the void fraction depending on the vessel height. Although these codes contain a one-dimensional model they allow the description of complex geometries due to the detailed nodalization of the considered apparatus. In chemical reactors, however, the venting process is not only influenced by the flashing behaviour but additionally by the running chemical reaction in the vessel. Therefore the codes used for modelling have to consider the kinetics of the chemical reaction. Further multi-component systems and dissolving processes have to be regarded. In order to preduct the fluid- and thermodynamic process it could be helpful to use 3-dimensional codes in combination with the one-dimensional codes as used in nuclear industry to get a more detailed describtion of the running processes. (orig./HP)

  8. 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.)

  9. A comparison of elastic-plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Anderson, C.A.; Smith, P.D.

    1979-01-01

    Numerical prediction of the behavior of prestressed concrete reactor vessels (PCRVs) under static, dynamic and long term loadings is complicated by the currently ill-defined behavior of concrete under stress and the three-dimensional nature of PCRVs. Which constitutive model most closely approximates the behavior of concrete in PCRVs under load has not yet been decided. Many equations for accurately modeling the three-dimensional behavior of PCRVs tax the capability of a most up-to-date computing system. The main purpose of this paper is to compare the characteristics of two constitutive models which have been proposed for concrete, variable modulus cracking model and elastic-plastic model. Moreover, the behavior of typical concrete structures was compared, the materials of which obey these constitutive laws. The response to internal pressure of PCRV structure, the constitutive models for concrete, the test problems using a thick-walled concrete ring and a rectangular concrete plate, and the analysis of an axisymmetric concrete pressure vessel PV-26 using the variable modulus cracking model of the ADINA code are explained. The variable modulus cracking model can predict the behavior of reinforced concrete structures well into the range of nonlinear behavior. (Kako, I.)

  10. 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

  11. Thermomechanical behavior of refractory-concrete-lined vessels

    Energy Technology Data Exchange (ETDEWEB)

    Tseng, T.M.

    1982-08-01

    A constitutive model applicable to dense and insulating refractory concretes used for coal gasification vessel linings is developed. Experiments were performed on refractory concrete plate specimens subjected to biaxial compression at various temperatures. Based on the test results, an orthotropic model is then developed. A temperature-dependent creep model based on the concept of thermorheologically simple material is also developed. The creep strain consists of a transient delayed elastic component and a nonlinear flow component. The developed constitutive model has been incorporated in a three-dimensional finite-element program to perform thermal stress analysis. To determine the temperature distributions as input to the stress analysis, a predictive model is developed for transient heat transfer analysis of structures with layered heterogeneous media. To allow for the effects of high conductivity of the media, an effective conductivity model is developed. Verification of the developed thermomechanical analysis is made through simulation runs for large-scale experimental gasifier vessels. The model is then used to perform parameter studies. The effect of heating rate, cooling rate, hold periods, shrinkage value, and thickness of the lining are studied. Design recommendations are made to improve the mechanical reliability of refractory concrete lined coal gasification vessels.

  12. 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.

  13. Chloride Ingress into Concrete under Water Pressure

    DEFF Research Database (Denmark)

    Lund, Mia Schou; Sander, Lotte Braad; Grelk, Bent

    2011-01-01

    The chloride ingress into concrete under water pressures of 100 kPa and 800 kPa have been investigated by experiments. The specimens were exposed to a 10% NaCl solution and water mixture. For the concrete having w/c = 0.35 the experimental results show the chloride diffusion coefficient at 800 k......Pa (~8 atm.) is 12 times greater than at 100 kPa (~1 atm.). For w/c = 0.45 and w/c = 0.55 the chloride diffusion coefficients are 7 and 3 times greater. This means that a change in pressure highly influences the chloride ingress into the concrete and thereby the life length models for concrete structures....

  14. Capacity of Prestressed Concrete Containment Vessels with Prestressing Loss

    International Nuclear Information System (INIS)

    SMITH, JEFFREY A.

    2001-01-01

    Reduced prestressing and degradation of prestressing tendons in concrete containment vessels were investigated using finite element analysis of a typical prestressed containment vessel. The containment was analyzed during a loss of coolant accident (LOCA) with varying levels of prestress loss and with reduced tendon area. It was found that when selected hoop prestressing tendons were completely removed (as if broken) or when the area of selected hoop tendons was reduced, there was a significant impact on the ultimate capacity of the containment vessel. However, when selected hoop prestressing tendons remained, but with complete loss of prestressing, the predicted ultimate capacity was not significantly affected for this specific loss of coolant accident. Concrete cracking occurred at much lower levels for all cases. For cases where selected vertical tendons were analyzed with reduced prestressing or degradation of the tendons, there also was not a significant impact on the ultimate load carrying capacity for the specific accident analyzed. For other loading scenarios (such as seismic loading) the loss of hoop prestressing with the tendons remaining could be more significant on the ultimate capacity of the containment vessel than found for the accident analyzed. A combination of loss of prestressing and degradation of the vertical tendons could also be more critical during other loading scenarios

  15. 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.

  16. Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1981-01-01

    The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed

  17. Liquid Nitrogen Subcooler Pressure Vessel Engineering Note

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, R.; /Fermilab

    1997-04-24

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

  18. 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

  19. Blood vessels, circulation and blood pressure.

    Science.gov (United States)

    Hendry, Charles; Farley, Alistair; McLafferty, Ella

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

  20. 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

  1. 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

  2. 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...

  3. 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...

  4. 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...

  5. 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...

  6. 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.

  7. Thermal effects, creep and nonlinear responde of concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1978-01-01

    A new mathematical model for prediction of pore pressure and moisture transfer in concrete heated well beyond 100 0 C is outlined. The salient features of the model are:(1) the hypothesis taht the pore space available to capillary water grows with increasing temperature as well as increasing pressure in excess of saturation pressure, and (2) the hypothesis that moisture permeability increases by two orders of magnitude when passing 100 0 C. Permaability below 100 0 C is controlled by migration of adsorbed water through gel-pore sized necks on passages through the material; these necks are lost above 100 0 C and viscosity then governs. The driving force of moisture transfer may be considered as the gradient of pore pressure, which is defined as pressure of vapor rather than liquid water if concrete is not saturated. Thermodynamic properties of water may be used to determine sorption isotherms in saturated concrete. The theory is the necessary first step in rationally predicting thermal stresses and deformations, and assessing the danger of explosive spalling. However, analysis of creep and nonlinear triaxial behavior is also needed for this purpose. A brief review of recent achievements in these subjects is also given. (Author)

  8. 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)

  9. 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.

  10. 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

  11. 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)

  12. 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

  13. 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

  14. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1983-01-01

    This construction of a container, which is pressure-relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 900 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW)

  15. Pressure vessel for gaseous media

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1977-01-01

    This construction of a container, which is pressure relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 90 0 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW) 891 RW [de

  16. Design and analysis of concrete reactor vessels: New developments, problems and trends

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1984-01-01

    This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance - areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of the problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep. (orig.)

  17. 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)

  18. 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.

  19. Preliminary analysis of a 1:4 scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Luk, V.K.; Hessheimer, M.F.

    1997-01-01

    Sandia National Laboratories is conducting a research program to investigate the integrity of nuclear containment structures. As part of the program Sandia will construct an instrumented 1:4 scale model of a prestressed concrete containment vessel (PCCV) for pressurized water reactors (PWR), which will be pressure tested up to its ultimate capacity. One of the key program objectives is to develop validated methods to predict the structural performance of containment vessels when subjected to beyond design basis loadings. Analytical prediction of structural performance requires a stepwise, systematic approach that addresses all potential failure modes. The analysis effort includes two and three-dimensional nonlinear finite element analyses of the PCCV test model to evaluate its structural performance under very high internal pressurization. Such analyses have been performed using the nonlinear concrete constitutive model, ANACAP-U, in conjunction with the ABAQUS general purpose finite element code. The analysis effort is carried out in three phases: preliminary analysis; pretest prediction; and post-test data interpretation and analysis evaluation. The preliminary analysis phase serves to provide instrumentation support and identify candidate failure modes. The associated tasks include the preliminary prediction of failure pressure and probable failure locations and the development of models to be used in the detailed failure analyses. This paper describes the modeling approaches and some of the results obtained in the first phase of the analysis effort

  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. 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

  2. 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

  3. Plan on test to failure of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.; Umeki, K.; Nagata, K.; Soejima, M.; Yamaura, Y.; Costello, J.F.; Riesemann, W.A. von.; Parks, M.B.; Horschel, D.S.

    1992-01-01

    A summary of the plans to test a prestressed concrete containment vessel (PCCV) model to failure is provided in this paper. The test will be conducted as a part of a joint research program between the Nuclear Power Engineering Corporation (NUPEC), the United States Nuclear Regulatory Commission (NRC), and Sandia National Laboratories (SNL). The containment model will be a scaled representation of a PCCV for a pressurized water reactor (PWR). During the test, the model will be slowly pressurized internally until failure of the containment pressure boundary occurs. The objectives of the test are to measure the failure pressure, to observe the mode of failure, and to record the containment structural response up to failure. Pre- and posttest analyses will be conducted to forecast and evaluate the test results. Based on these results, a validated method for evaluating the structural behavior of an actual PWR PCCV will be developed. The concepts to design the PCCV model are also described in the paper

  4. Pretest round robin analysis of 1:4-scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Hessheimer, M.F.; Luk, V.K.; Klamerus, E.W.; Shibata, S.; Mitsugi, S.; Costello, J.F.

    2001-01-01

    The work reported herein represents, arguably, the state of the art in the numerical simulation of the response of a prestressed concrete containment vessel (PCCV) model to pressure loads up to failure. A significant expenditure of time and money on the part of the sponsors, contractors, and Round Robin participants was required to meet the objectives. While it is difficult to summarize the results of this extraordinary effort in a few paragraphs, the following observations are offered for the reader's consideration: almost half the participants used ABAQUS as the primary computational tool for performing the pretest analyses. The other participants used a variety of codes, most of which were developed ''in house''. (author)

  5. Shear strength of end slabs of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Cheung, K.C.; Gotschall, H.L.; Liu, T.C.

    1975-01-01

    Prestressed concrete reactor vessels (PCRV's) have been adopted for primary containments in most large high-temperature gas-cooled reactor installations. The most common configuration for PCRVs is a right-vertical cylinder with thick end slabs. In order to assess the integrity of a PCRV it is necessary to predict the ultimate strength of the end slabs. The complexity of the basic mechanism of shear failure in the PCRV end slabs has thus far prohibited the development of a completely analytical solution. However, many experimental investigations of PCRV end slabs have been conducted over the past decade. This information makes it possible to establish empirical formulae for the ultimate strength of PCRV end slabs. The basis and development of an empirical shear-flexure interaction expression is presented. (Auth.)

  6. Stress analysis of pierced regions for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1983-01-01

    The analysis of prestressed concrete vessels for nuclear reactors and, in particular, the design of the perforated end-caps increasingly calls for better analytical techniques. A method is proposed by which the distribution of stress in the perforated end-caps may be defined, assuming constant elevated temperature and radial prestress. A special feature of the proposed analysis is the method by which an elastic solution can be modified to take account of the time-dependent effects such as creep and shrinkage at ambient and elevated temperatures. The technique was used to analyse slab models. It was shown that the proposed method estimates the deformational behaviour of the models with an acceptable degree of accuracy. The distribution of stress as a result of thermal creep predicted by the method also agreed well with the cracking pattern of the models. (U.K.)

  7. 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

  8. 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.

  9. 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)

  10. 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.

  11. Nonlinear analysis of pre-stressed concrete containment vessel (PCCV) using the damage plasticity model

    Energy Technology Data Exchange (ETDEWEB)

    Shokoohfar, Ahmad; Rahai, Alireza, E-mail: rahai@aut.ac.ir

    2016-03-15

    Highlights: • This paper describes nonlinear analyses of a 1:4 scale model of a (PCCV). • Coupled temp-disp. analysis and concrete damage plasticity are considered. • Temperature has limited effects on correct failure mode estimation. • Higher pre-stressing forces have limited effects on ultimate radial displacements. • Anchorage details of liner plates leads to prediction of correct failure mode. - Abstract: This paper describes the nonlinear analyses of a 1:4 scale model of a pre-stressed concrete containment vessel (PCCV). The analyses are performed under pressure and high temperature effects with considering anchorage details of liner plate. The temperature-time history of the model test is considered as an input boundary condition in the coupled temp-displacement analysis. The constitutive model developed by Chang and Mander (1994) is adopted in the model as the basis for the concrete stress–strain relation. To trace the crack pattern of the PCCV concrete faces, the concrete damage plasticity model is applied. This study includes the results of the thermal and mechanical behaviors of the PCCV subject to temperature loading and internal pressure at the same time. The test results are compared with the analysis results. The analysis results show that the temperature has little impact on the ultimate pressure capacity of the PCCV. To simulate the exact failure mode of the PCCV, the anchorage details of the liner plates around openings should be maintained in the analytical models. Also the failure mode of the PCCV structure hasn’t influenced by hoop tendons pre-stressing force variations.

  12. 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.)

  13. 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.)

  14. 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.

  15. Safety of steel vessel Magnox pressure circuits

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  16. Time varying stress in ligaments of perforated plates with reference to prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1978-01-01

    The work described herein relates to the prediction of stresses in materials which exhibit time varying strains with particular reference to the ligaments of perforated circular concrete slabs, subjected to long-term radial prestress and uniform elevated temperature. The perforations are reinforced with steel liners and arranged in a square central lattice symmetrical about two orthogonal axes. Special reference is made to the distribution of stress in the standpipe region of prestressed concrete cylindrical pressure or containment vessels for gas cooled reactors. In order to assess the stress distribution around the perforated zone of a circular slab, a method of analysis was developed by the author, based on the ''Equivalent Elastic Modulus'' of the perforated zone and the ''Effective Modulus Method'', utilizing experimental data obtained from tests performed on model specimens. The object of this paper is to extend the above method of analysis into the perforated region, and assess the long-term stresses in the ligaments. The proposed method is accomplished by an application of the Finite Element Method for the elastic plane stress case. Comparisons of experimental results and theoretical predictions by the proposed method, and other analytical methods are made for a series of perforated concrete slabs subjected to radial in-plane loading: 10,342 kN/m 2 (1,5000 psi), and uniform elevated temperature of 80 0 C. The investigation, though in general terms, could be applied to the perforated region of cylindrical pressure vessels for nuclear reactors. Finally the paper describes briefly in Appendix 3 a direct solution procedure for calculating time dependent stresses in concrete structures based on the principles of variational calculus. Analytical predictions obtained by the proposed method which is a step-by-step analysis, are compared with the variational principle method. (author)

  17. Ultimate load design and testing of a cylindrical prestressed concrete vessel

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1982-01-01

    The object of this research was to design, construct and test to failure a prestressed concrete pressure vessel model that could be used to investigate the behavior of a full scale structure underworking and ultimate load. The properties and the design of the model was based generally on full scale vessels already constructed to house the nuclear reactors used in atomic power stations. To design the model the ultimate load approach was adopted throughout. All load factors associated with the prestressing have been defined and kept to a minimum in order that the vessel's behavior may be predicted. The tests on the vessel were carried out first on the elastic range to observe its behavior at working load and then at the ultimate range to observe the modes of failure and compare the actual results in both cases with the predicted values. Although full agreement between observed results and predicted values was not obtained, the conclusions drawn from the study were useful for the design of full scale vessels. (author)

  18. Shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Reins, J.D.; Quiros, J.L. Jr.; Schnobrich, W.C.; Sozen, M.A.

    1976-07-01

    The report summarizes the experimental and part of the analytical work carried out in connection with an investigation of the structural strength of prestressed concrete reactor vessels. The project is part of the Prestressed Concrete Reactor Vessel Program of the Oak Ridge National Laboratory sponsored by ERDA. The objective of the current phase of the work is to develop procedures to determine the shear strength of flat end slabs of reactor vessels with penetrations

  19. 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...

  20. 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...

  1. 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.)

  2. 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.

  3. 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)

  4. 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

  5. Seismic analysis of a reinforced concrete containment vessel model

    International Nuclear Information System (INIS)

    Randy, James J.; Cherry, Jeffery L.; Rashid, Yusef R.; Chokshi, Nilesh

    2000-01-01

    Pre-and post-test analytical predictions of the dynamic behavior of a 1:10 scale model Reinforced Concrete Containment Vessel are presented. This model, designed and constructed by the Nuclear Power Engineering Corp., was subjected to seismic simulation tests using the high-performance shaking table at the Tadotsu Engineering Laboratory in Japan. A group of tests representing design-level and beyond-design-level ground motions were first conducted to verify design safety margins. These were followed by a series of tests in which progressively larger base motions were applied until structural failure was induced. The analysis was performed by ANATECH Corp. and Sandia National Laboratories for the US Nuclear Regulatory Commission, employing state-of-the-art finite-element software specifically developed for concrete structures. Three-dimensional time-history analyses were performed, first as pre-test blind predictions to evaluate the general capabilities of the analytical methods, and second as post-test validation of the methods and interpretation of the test result. The input data consisted of acceleration time histories for the horizontal, vertical and rotational (rocking) components, as measured by accelerometers mounted on the structure's basemat. The response data consisted of acceleration and displacement records for various points on the structure, as well as time-history records of strain gages mounted on the reinforcement. This paper reports on work in progress and presents pre-test predictions and post-test comparisons to measured data for tests simulating maximum design basis and extreme design basis earthquakes. The pre-test analyses predict the failure earthquake of the test structure to have an energy level in the range of four to five times the energy level of the safe shutdown earthquake. The post-test calculations completed so far show good agreement with measured data

  6. 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.)

  7. 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

  8. 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

  9. 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)

  10. 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

  11. Analyses and testing of model prestressed concrete reactor vessels with built-in planes of weakness

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Fleischer, C.C.

    1990-01-01

    This paper describes the design, construction, analyses and testing of two small scale, single cavity prestressed concrete reactor vessel models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. This work was carried out to extend a previous study which had suggested the likely feasibility of constructing regions of prestressed concrete reactor vessels and biological shields, which become activated, using easily removable blocks, separated by a suitable membrane. The paper describes the results obtained and concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of prestressed concrete reactor vessels, biological shields and similar types of structure. (author)

  12. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  13. 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)

  14. 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.

  15. The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

    International Nuclear Information System (INIS)

    Bolvin, M.

    1981-01-01

    This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

  16. 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)

  17. 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

  18. 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.)

  19. 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

  20. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  1. 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.

  2. 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.

  3. 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

  4. Pressurized wet digestion in open vessels (T11)

    International Nuclear Information System (INIS)

    Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

    2002-01-01

    Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

  5. 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.

  6. 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

  7. State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability

    International Nuclear Information System (INIS)

    Bonnet, Jean-Michel; Cranga, Michel; Vola, Didier; Marchetto, Cathy; Kissane, Martin; ); Robledo, Fernando; Farmer, Mitchel T.; Spengler, Claus; Basu, Sudhamay; Atkhen, Kresna; Fargette, Andre; Fisher, Manfred; Foit, Jerzi; Hotta, Akitoshi; Morita, Akinobu; Journeau, Christophe; Moiseenko, Evgeny; Polidoro, Franco; Zhou, Quan

    2017-01-01

    Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures

  8. Pressure-tension test for assessing fatigue in concrete

    Science.gov (United States)

    Soleimani, Sayed M.; Boyd, Andrew J.; Komar, Andrew J. K.

    2017-04-01

    In a pressure-tension test, a cylindrical concrete specimen is inserted into a cylindrical steel jacket, with a rubber ``O'' ring seal at each end to prevent gas leakage. Gas pressure is then applied to the curved surface of the concrete cylinder, leaving the ends free. As the gas pressure is increased, the specimen eventually fractures across a single plane transverse to the axis of the cylinder. The gas pressure at fracture may then be considered as the tensile strength of the concrete. In this study, the pressure-tension test is used to study fatigue in concrete. A total of 22 standard concrete cylinders (100 mm × 200 mm) were tested. Both dry and wet specimens have been studied. Low-cycle loading, which involves the application of a few load cycles at high stress levels - such as a concrete structure under earthquake load - has been used in this study. It was found that the concrete specimens in a low-cycle loading fail after only a few cycles of loading and interestingly at a stress level lower than the maximum value applied in the cyclic loading. In addition, non-destructive testing (NDT) was performed to determine the progressive damage due to tensile load in concrete cylinders using Ultrasonic Pulse Velocity (UPV). It was found that UPV can be used to evaluate the damage in concrete even after the application of a very low-level of tensile stress - as low as 10% of its tensile strength.

  9. Pre-stressed concrete reactor vessel with built-in planes of weakness

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Fleischer, C.C.

    1989-01-01

    This report describes a study that has been carried out to extend previous work to investigate the feasibility of constructing regions of pre-stressed concrete reactor vessels (PCRV) and biological shields which become activated using easily removable blocks, separated by a suitable membrane. The previous study concluded that, from preliminary analyses, such a concept appeared feasible and recommended that further work should be done. The present study was therefore commissioned to carry out more detailed analytical work and to complement this with the design, construction and pressure testing of two small-scale, single-cavity PCRV models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. The report describes the analyses, the model design, construction and testing, and presents relevant results. It concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of PCRV, biological shields and similar types of structure

  10. 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

  11. Reliability based approach for the ageing management of prestressed concrete containment vessel

    International Nuclear Information System (INIS)

    Heinfling, G.; Courtois, A.; Viallet, E.

    2005-01-01

    The containment of the French 1300/1450 MWe pressurized water reactors is ensured by two concrete vessels. The inner containment is biaxially prestressed so that it remains in compression under the pressure and temperature loading associated with a LOCA. It's design is based on hypotheses concerning creep and shrinkage induced loss of prestress. These phenomena must be accurately monitored and their evolution must be accurately predicted in order to estimate the capability of the structure to undergo accidental conditions in the future during its whole industrial lifetime. The prediction of the further delayed behaviour of the containment vessel must be able to account for the previously monitored behaviour, the uncertainties associated to the measurements and the uncertainties associated to the model parameters. A reliability based method has been developed by EDF in that aim. This method allows to obtain an updated statistical distribution of the delayed strain. The modelling of the delayed behaviour is based on a physical model which accounts for the decomposition of creep and shrinkage physical mechanisms. The uncertainties on the corresponding physical parameters are accounted for by coupling this model with a classical probabilistic method. The updating of the initial prediction on the base of the measured delayed strains is performed trough a bayesian technique. (authors)

  12. 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)

  13. Measured Prestress Loss of over 20-Year-Old Prestressed Concrete Containment Vessels

    International Nuclear Information System (INIS)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil

    2010-01-01

    Most nuclear reactors, both in Korea and worldwide, are enclosed by a prestressed concrete containment vessels(PCCVs). The containment wall is approximately 1 m thick and is prestressed in two directions by large prestressing tendons. The main purpose of the containment is to maintain the structural integrity of the containment in the event of a major internal accident. The main accidental scenario, which the containment is designed to withstand, is a so-called loss of coolant accident (LOCA). A LOCA is initiated by a pipe rupture in the cooling system, discharging hot steam into the containment. The escape of steam increases both the temperature and pressure inside the containment. The increased internal pressure arising from a LOCA is referred to as the design pressure. The prestressing system is designed to counterbalance the tensile forces arising from the design pressure. The status of the containment is gradually changed due to environmental factors and by alterations in the micro structure of the material. The prestress will be reduced due to shrinkage and creep in the concrete and relaxation in the tendons. The corrosion protection of tendons are for Korean containments arranged in two different ways, either by cement grouting (bonded tendons) or e.g. by grease injection (unbonded tendons). The major advantage using unbonded tendons is the possibilities of assessing their status (e.g. prestress losses or corrosion damages) which is not possible using bonded tendons. Both bonded and unbonded tendons are used worldwide. For example in the U.S. almost all tendons are unbonded, whereas in France almost all tendons are bonded. For Korean reactor containments with unbonded tendons (14 containments) the tendon force is monitored at regular in-service inspections. The power plant Wolsung in Korea has bonded tendons and several prestressed concrete beams were constructed with the single purpose to follow up the prestress losses. The remaining tendon forces in some

  14. Plastic collapse pressure of cylindrical vessels containing longitudinal surface cracks

    Energy Technology Data Exchange (ETDEWEB)

    Zarrabi, K. [New South Wales Univ., Sydney, NSW (Australia). Sch. of Mech. and Mfg. Eng.; Zhang, H. [New South Wales Univ., Sydney, NSW (Australia). Sch. of Mech. and Mfg. Eng.; Nhim, K. [New South Wales Univ., Sydney, NSW (Australia). Sch. of Mech. and Mfg. Eng.

    1997-05-01

    Based on nonlinear finite element analysis, the plastic collapse pressures of cylindrical vessels with longitudinal surface cracks are computed. A general formula of plastic collapse pressure of such structures are given and compared with the literature solutions. The results of the present study could be applied for the integrity assessments, failure analyses, remanent life assessment, and licence extensions of the vessels. (orig.)

  15. 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...

  16. Experimental verification of creep analyses for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Abe, H.; Ohnuma, H.

    1977-01-01

    The authors proposed a new method of creep analysis based on the theory of strain hardening, which assumes that accumulated creep at a given time influences the creep after that. This method was applied to calculate step-by-step the behaviors of uniaxial creep of concrete under variable temperatures and stresses, creep in reinforced concrete specimens and the behaviors of prestressed concrete beams under themal gradients. The experimental and calculated results agreed fairly well. Further, this method was incorporated in the finite element creep analysis for the prestressed concrete hollow cylinder and the full scale model. The calculated strain changes with time pursued closely those obtained by experiments. The above led to the conclusion that from the viewpoint of both accuracy and computation time the strain hardening method proposed by the authors may be judged advantageous for practical usages

  17. 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

  18. 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.)

  19. Effects of no stiffness inside unbonded tendon ducts on the behavior of prestressd concrete containment vessels

    International Nuclear Information System (INIS)

    Noh, Sang Hoon; Kwak, Hyo Gyong; Jung, Rae Young; Noh, Sang Hoon

    2016-01-01

    The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading

  20. Effects of no stiffness inside unbonded tendon ducts on the behavior of prestressd concrete containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sang Hoon; Kwak, Hyo Gyong [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jung, Rae Young; Noh, Sang Hoon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-06-15

    The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading.

  1. Posttest analysis of a 1:4-scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Hessheimer, M.F.

    2003-01-01

    The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, co-sponsored a Cooperative Containment Research Program at Sandia National Laboratories (SNL) in Albuquerque, New Mexico. As part of the program, a prestressed concrete containment vessel (PCCV) model was subjected to a series of overpressurization tests at SNL beginning in July 2000 and culminating in a functional failure mode or Limit State Test (LST) in September 2000 and a Structural Failure Mode Test (SFMT) in November 2001. The PCCV model, uniformly scaled at 1:4, is representative of the containment structure of an actual Pressurized Water Reactor (PWR) plant (OHI-3) in Japan. The objectives of the pressurization tests were to obtain measurement of the structural response to pressure loading beyond design basis accident in order to validate analytical modeling, to find pressure capacity of the model, and to observe its failure mechanisms. This paper compares results of pretest analytical studies of the PCCV model to the PCCV high pressure test measurements and describes results of post-test analytical studies. These analyses have been performed by ANATECH Corp. under contract with Sandia National Laboratories. The post-test analysis represents the third phase of a comprehensive PCCV analysis effort. The first phase consisted of preliminary analyses to determine what finite element models would be necessary for the pretest prediction analyses, and the second phase consisted of the pretest prediction analyses. The principal objectives of the post-test analyses were: (1) to provide insights to improve the analytical methods for predicting the structural response and failure modes of a prestressed concrete containment, and (2) to evaluate by analysis any phenomena or failure mode observed during the test that had not been explicitly predicted by analysis. In addition to summarizing comparisons between measured

  2. 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

  3. Systems aspects of reactor pressure vessel integrity: Executive report

    International Nuclear Information System (INIS)

    Ray, N.K.; Beck, R.K.; Wallace, I.W.; Grigsby, J.M.; Sloane, B.D.; Bishop, B.A.

    1993-11-01

    Reactor vessel integrity is affected by the following key issues: Operational heatup and cooldown pressure-temperature limit curves and the associated occurrence of low temperature over pressurization (LTOP); pressurized thermal shock (PTS) events that could lead to crack initiation in the vessel, and irradiation effects, including low upper shelf fracture toughness. Electric Power Research Institute's Embrittlement Management Program is underway to develop methods in assessing reactor vessel integrity issues due to radiation embrittlement. Reactor vessel integrity may be impacted by different aspects including some degradation mechanisms. One of those aspects are the plant system changes associated with the reactor vessel. This report uses the results from various probabilistic analyses performed by Westinghouse and other organizations, to identify those plant systems which are most important to the reactor vessel embrittlement issues

  4. 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.

  5. 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

  6. Experimental investigations concerning the suitability of channel systems for liner leak detection and drainage of a prestressed concrete vessel

    International Nuclear Information System (INIS)

    Nickel, M.; Breitbach, G.; Altes, J.; Escherich, K.H.; Wolters, J.

    1985-02-01

    The iternal surfaces of prestressed concrete pressure vessels are fitted with a steel liner to preserve the gas tightness of the primary circuit. Because of the high quality manufacture and the loading conditions a linear failure can be practically excluded. However, if it is hypothetically assumed, that a leak develops during reactor operation, it may be difficult to determine the position of the leak, because the linear area is very large. For tightness surveillance and for venting channel systems installed in close proximity to the linear are suitable. The suitability of such channels for leak detection, localisation and venting was investigated experimentally. A concrete wall (length 2.5 m, height 2.0 m, thickness 0.5 m) was constructed, covered on one side with a steel liner. Behind the liner two different channel systems have been installed. For the simulation of leaks holes were drilled into the liner. The experimental programm contained the following measurements: determination of gas flow rates into the different leaks, distribution of leakage gas over the array of channels and determination of pressures into the concrete and immediately behind the liner. The experiments have shown, that channel arrays immediately adjacent to the liner are the most suitable systems for localisation and controlled exhaust of leakage gas. The suitability decreases, if the channels are set into the concrete somewhat distant from the liner. (orig.) [de

  7. The permeability of concrete for reactor containment vessels

    International Nuclear Information System (INIS)

    Mills, R.H.

    1983-07-01

    Review of the literature pertaining to water, water vapour and gas transmission through concrete revealed conflicting views on the mechanisms involved and the influence of mix design parameters such as initial porosities and water/cement ratio. Consideration of the effects of ageing and of construction defects in field concrete were totally neglected in published work. Permeability data from three published papers were compared with permeability calculated according to Powers. The ratio of calculated to observed permeability varied from 40 x 10 -3 to 860 x 10 -3 for one group: from 0.17 x 10 3 to 8.6 x 10 3 in the second; and from 24 x 10 3 to 142 x 10 3 for the third. There were therefore wide discrepancies within each group of data and between groups. A bibliography was prepared and an exploratory experimental programme was mounted to determine the relative importance of key parameters such as cement type, porosity and water/cement ratio. Contrary to frequently cited references it was found that permeability of concrete was not significantly influenced by water/cement ratio when the starting porosity was constant. If water/cement ratio was held constant, however, the permeability was strongly influenced by starting porosity. It was also found that with constant water/cement ratio permeability increased with cement content. The value of fly ash and blast furnace slag in partial substitution for Portland cement is neglected in the literature but it is important since such substitutions alleviate alkali-silicate reactions. Permeability of concrete was significantly decreased by partial substitution of Portland cement with fly ash but there was no benefit in the use of blast furnace slag

  8. 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

  9. 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...

  10. 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

  11. 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

  12. 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.

  13. 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 ...

  14. Heavy wall pressure vessels for energy systems

    International Nuclear Information System (INIS)

    Canonico, D.A.

    Modifications of steels currently accepted in the Code appear to provide improved mechanical properties. These steels may permit the fabrication of larger diameter vessels with thinner section sizes and improved reliability and integrity. Adapting current specifications should expedite Code approval. Finally the challenge of improving welding procedures and adapting processes for field applications will result in higher quality weldments

  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. Evaporation drying takes place from the unsealed end of the specimen. A drying front moves into the concrete and considerable weight loss is recorded as moisture escapes to the atmosphere. The rate of movement of the drying front is slower than that of the hot front and is proportional to the temperature difference between the top of the specimen and the surrounding atmosphere. In the shrinkage specimen, values of transverse and longitudinal shrinkage reflect the water content results. The specimen indicates that shrinkage occurs in a concrete pressure vessel, in the regions where moisture is lost. The restraint of the mass of concrete surrounding these regions sets up a three dimensional state of internal tensile stress. The areas into which the moisture migrates tend to swell, creating an internal stress situation, which is this

  16. Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

    International Nuclear Information System (INIS)

    Choun, Young Sun; Park, Jun Hee

    2015-01-01

    Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers

  17. Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Young Sun; Park, Jun Hee [Integrated Safety Assessment Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete containment vessel (PCCV) were investigated in this study. For a comparative evaluation between the shear performances of structural walls constructed with conventional concrete, steel fiber reinforced concrete, and polyamide fiber reinforced concrete, cyclic tests for wall specimens were conducted and hysteretic models were derived. The shear resisting capacity of a PCCV constructed with fiber reinforced concrete can be improved considerably. When steel fiber reinforced concrete contains hooked steel fibers in a volume fraction of 1.0%, the maximum lateral displacement of a PCCV can be improved by > 50%, in comparison with that of a conventional PCCV. When polyamide fiber reinforced concrete contains polyamide fibers in a volume fraction of 1.5%, the maximum lateral displacement of a PCCV can be enhanced by ∼40%. In particular, the energy dissipation capacity in a fiber reinforced PCCV can be enhanced by > 200%. The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers.

  18. A system for the thermal insulation of a pre-stressed concrete vessel

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1975-01-01

    This invention concerns the thermal insulation of a pre-stressed concrete vessel for a pressurised water nuclear reactor, this vessel being fitted internally with a leak-proof metal lining. Two rings are placed at the lower and upper parts of the vessel respectively. The upper ring is closed with a cover. These rings differ in diameter, are fitted with a metal insulating and mark the limits of a chamber between the vaporisable fluid and the internal wall of the vessel. This chamber is filled with a fluid in the liquid phase up to the liquid/vapor interface level of the fluid and with a gas above that level, the covering of the rings forming a cold fluid liquid seal. Each ring is supported by the vessel. Leak-proof components take up the radial expansion of the rings [fr

  19. Investigation of impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-10-15

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

  20. Stress analysis of large pressure vessels

    International Nuclear Information System (INIS)

    Jeanes, P.; King, S.; Wright, M.B.

    1981-01-01

    Five spherical vessels were investigated in this study. Each model had approximately 2,500 20-node solid elements and 12,000 grid points. Three finite element method computer programs were used, MSC/NASTRAN, BERSAFE and FEMALE. The general approach to the problem, the approach to structural modelling, including the shell and nozzles, the stress analysis and the processing of the results are summarized. (U.K.)

  1. 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...

  2. Transportable, small high-pressure preservation vessel for cells

    International Nuclear Information System (INIS)

    Kamimura, N; Sotome, S; Shimizu, A; Nakajima, K; Yoshimura, Y

    2010-01-01

    We have previously reported that the survival rate of astrocytes increases under high-pressure conditions at 4 0 C. However, pressure vessels generally have numerous problems for use in cell preservation and transportation: (1) they cannot be readily separated from the pressurizing pump in the pressurized state; (2) they are typically heavy and expensive due the use of materials such as stainless steel; and (3) it is difficult to regulate pressurization rate with hand pumps. Therefore, we developed a transportable high-pressure system suitable for cell preservation under high-pressure conditions. This high-pressure vessel has the following characteristics: (1) it can be easily separated from the pressurizing pump due to the use of a cock-type stop valve; (2) it is small and compact, is made of PEEK and weighs less than 200 g; and (3) pressurization rate is regulated by an electric pump instead of a hand pump. Using this transportable high-pressure vessel for cell preservation, we found that astrocytes can survive for 4 days at 1.6 MPa and 4 0 C.

  3. 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.

  4. Development of ultrasonic testing technique with the large transducer to inspect the containment vessel plates of nuclear power plant embedded in concrete

    International Nuclear Information System (INIS)

    Ishida, Hitoshi; Kurozumi, Yasuo; Kaneshima, Yoshiari

    2004-01-01

    The containment vessel plates embedded in concrete on Pressurized Water Reactors are inaccessible to inspect directly. Therefore, it is advisable to prepare inspection technology to detect existence and a location of corrosion on the embedded plates indirectly. In order to establish ultrasonic testing technique to be able to inspect the containment vessel plates embedded in concrete widely at the accessible point, experiments to detect artificial hollows simulating corrosion on a surface of a carbon steel plate mock-up covered with concrete simulating the embedded containment vessel plates were carried out with newly made ultrasonic transducers. We made newly low frequency (0.3 MHz and 0.5 MHz) surface shear horizontal (SH) wave transducers combined with three large active elements, which were equivalent to a 120mm width element. As a result of the experiments, the surface SH transducers could detect clearly the echo from the hollows with a depth of 9.5 mm and 19 mm at a distance of 1500mm from the transducers on the surface of the mock-up covered with concrete. Therefore, we evaluate that it is possible to detect the defects such as corrosion on the plates embedded in concrete with the newly made low frequency surface SH transducers with large elements. (author)

  5. 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)

  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. 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.

  9. 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.)

  10. Preliminary investigation on the suitablity of using fiber reinforced concrete in the construction of a hazardous waste disposal vessel

    International Nuclear Information System (INIS)

    Ramey, M.R.; Daie-e, G.

    1988-07-01

    There are certain hazardous wastes that must be contained in an extremely secure vessel for transportation and disposal. The vessel, among other things, must be able to withstand relatively large impacts without rupturing. Such containment vessels therefore must be able to absorb substantial amounts of energy during an impact and still perform their function. One of the impacts that the vessel must withstand is a 30-foot fall onto an unyielding surface. For some disposal scenarios it is proposed to encase the waste in a steel enclosure which is to be surrounded by a thick layer of concrete which, in turn, is encased by a relatively thin steel shell. Tests on concrete in compression and flexure, including static, dynamic and impact tests, have shown that low modulus concretes tend to behave in a less brittle manner than higher modulus concretes. Tests also show that fiber reinforced concretes have significantly greater ductility, crack propagation resistance and toughness than conventional concretes. Since it is known that concrete is a reasonably brittle material, it is necessary to do impact tests on sample containment structures consisting of thin-walled metal containers having closed ends which are filled with concrete, grout, or fiber reinforced concrete. This report presents the results of simple tests aimed at observing the behavior of sample containment structures subjected to impacts due to a fall from 30 feet. 8 figs., 4 tabs

  11. 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

  12. 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

  13. 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).

  14. 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).

  15. 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

  16. 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)

  17. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

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

    2006-01-01

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

  18. 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.)

  19. 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

  20. 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.

  1. Analysis and application of prestressed concrete reactor vessels for LMFBR containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Fistedis, S.H.; Bazant, Z.P.; Belytschko, T.B.

    1978-01-01

    An analytical model of a prestressed concrete reactor vessel (PCRV) for LMFBR and the associated finite element computer code, involving an explicit time integration procedure, is described. The model is axisymmetric and includes simulations of the tensile cracking of concrete, the reinforcement, and a prestressing capability. The tensile cracking of concrete and the steel reinforcement are both modeled as continuously distributed within the finite element. The stresses in the reinforcement and concrete are computed separately and combined to give an overall stress state of the composite material. Attention is given to the fact that cracks do not form instantaneously, but develop gradually. Thus, after crack initiation the normal stress is reduced to zero gradually as a function of time. Residual shear resistance of cracks due to aggregate interlock is also taken into account. Prestressing of the PCRV is modeled by special structural members which represent an averaged prestressing layer equivalent to an axisymmetric shell. The internal prestressing members are superimposed over the reinforced concrete body of the PCRV; they are permitted to stretch and slide in a predetermined path, simulating the actual tendons. The validity of the code is examined by comparison with experimental data. (Auth.)

  2. Influence of Wind Pressure on the Carbonation of Concrete

    Directory of Open Access Journals (Sweden)

    Dujian Zou

    2015-07-01

    Full Text Available Carbonation is one of the major deteriorations that accelerate steel corrosion in reinforced concrete structures. Many mathematical/numerical models of the carbonation process, primarily diffusion-reaction models, have been established to predict the carbonation depth. However, the mass transfer of carbon dioxide in porous concrete includes molecular diffusion and convection mass transfer. In particular, the convection mass transfer induced by pressure difference is called penetration mass transfer. This paper presents the influence of penetration mass transfer on the carbonation. A penetration-reaction carbonation model was constructed and validated by accelerated test results under high pressure. Then the characteristics of wind pressure on the carbonation were investigated through finite element analysis considering steady and fluctuating wind flows. The results indicate that the wind pressure on the surface of concrete buildings results in deeper carbonation depth than that just considering the diffusion of carbon dioxide. In addition, the influence of wind pressure on carbonation tends to increase significantly with carbonation depth.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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.)

  9. 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)

  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. 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...

  12. 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...

  13. 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...

  14. 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.

  15. 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

  16. Stress analysis in a non axisymmetric loaded reactor pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de; Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel

    1995-01-01

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author)

  17. 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.

  18. 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.)

  19. 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

  20. Improved Attachment in a Hybrid Inflatable Pressure Vessel

    Science.gov (United States)

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

    2010-01-01

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

  1. 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

  2. Mechanical probabilistic study of pressurized water reactor (PWR) vessels

    International Nuclear Information System (INIS)

    Venturini, V.; Persoz, M.; Debost-Eymard; Donore, A.M.; Meister, E.

    1999-01-01

    This paper describes the probabilistic study carried out on the vessel case under assumptions of occurrence of an underclad flaw and of an accidental transient of the pressurized thermal shock type (PTS). The estimation model of the failure criterion requires finite element computations. Therefore the probabilistic analysis is based on a coupling between Code Aster, the EDF FE software for thermomechanical analysis and PROBAN, a probabilistic analysis software (Det Norske Veritas). The vessel model is first described by focusing on the thermomechanical model (pseudo-1D and 2D axisymmetrical computations are considered and compared) then on the model used in the probabilistic analysis. The second part deals with the probabilistic analyses (computations of failure probabilities during the transient) carried out on the model for different cases and the comparison of the results. Eventually, the perspectives of this work, from both the computational point of view and the application to the vessel study, are outlined. (orig./DGE)

  3. 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...

  4. 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...

  5. 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)

  6. 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)

  7. 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.

  8. 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.

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

    CERN Document Server

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

    1998-01-01

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

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

    Science.gov (United States)

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

    2007-01-01

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

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  13. 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.

  14. Backstreaming of Impurity Gas Through a Leak in Pressurized Vessel

    CERN Document Server

    Dauvergne, J P

    1998-01-01

    The presence of a leak in a vessel containing pure gas can induce the contamination by atmospheric gas diffusing into the vessel. In order to avoid this, a gas which has to be kept pure also in presen ce of a leak is usually pressurized, to reduce the flow of contaminating gas through the leak owing to the molecular drag by the outstreaming pure gas. In this paper, a simple model calculation of ba ckstreaming based on the solution of the diffusion + drag equation in cylindrical coordinates is presented. It is shown that both the pressure difference and the dimension of the leak are critical in determining the contaminating flow, a maximum in the backstreaming flow appearing when the drag velocity of the outstreaming gas equals the diffusion velocity.

  15. 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.

  16. 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.

  17. 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)

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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)

  3. 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.

  4. Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

  5. Prestressed Concrete-Lined Pressure Tunnels : Towards Improved Safety and Economical Design

    NARCIS (Netherlands)

    Simanjuntak, Y.

    2015-01-01

    Pressure tunnels for hydropower plants are relatively expensive constructions, particularly when steel linings are used. Concrete linings can be economically attractive; however, their applicability is limited by the low tensile strength of concrete. Techniques to improve the bearing capacity of

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

    International Nuclear Information System (INIS)

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

    1994-10-01

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

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

    Science.gov (United States)

    2010-10-01

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

  8. 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...

  9. 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...

  10. 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...

  11. 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...

  12. 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...

  13. 77 FR 59408 - Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying...

    Science.gov (United States)

    2012-09-27

    ... Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII with respect to...; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying Liquefied Gases in Bulk AGENCY: Coast...-ENG Policy Letter 04-12, ``Alternative Pressure Relief Valve Settings on Vessels Carrying Liquefied...

  14. 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.)

  15. 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

  16. 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

  17. 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.

  18. 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.

  19. 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.)

  20. 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

  1. 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)

  2. 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.

  3. 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)

  4. 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.

  5. 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.

  6. Thermodynamic study on the in-vessel corium - Application to the corium/concrete interaction

    International Nuclear Information System (INIS)

    Quaini, Andrea

    2015-01-01

    During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore, depending on the considered scenario, the corium can be formed by a liquid phase or by two liquids, one metallic the other oxide. The objective of this thesis is the investigation of the thermodynamics of the prototypic in-vessel corium U-Pu-Zr- Fe-O. The approach used during the thesis is based on the CALPHAD method, which allows to obtain a thermodynamic model for this complex system starting from phase diagram and thermodynamic data. Heat treatments performed on the O-U-Zr system allowed to measure two tie-lines in the miscibility gap in the liquid phase at 2567 K. Furthermore, the liquidus temperatures of three Zr-enriched samples have been obtained by laser heating in collaboration with ITU. With the same laser heating technique, solidus temperatures have been obtained on the UO 2 -PuO 2 -ZrO 2 system. The influence of the reducing or oxidising on the melting behaviour of this system has been studied for the first time. The results show that the oxygen stoichiometry of these oxides strongly depends on the oxygen potential and on the metal composition of the samples. The miscibility gap in the liquid phase of the U-Zr-Fe-O system has been also observed. The whole set of experimental results with the literature data allowed to develop the thermodynamic model of the U-Pu-Zr-Fe-O system. Solidification path calculations have been performed for all the investigated samples to interpret the microstructures of the solidified samples. A good accordance has been obtained between

  7. Test of 6-inch-thick pressure vessels. Series 2. Intermediate test vessels V-3, V-4, and V-6

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Raftenberg, M.N.; Robinson, G.C.; Smith, J.E.

    1975-11-01

    The second series of intermediate vessel tests were crack initiation fracture tests of 6-in.-thick 39-in.-OD steel vessels with sharp surface flaws approximately 2 1 / 2 in. deep by 8 in. long in the longitudinal weld seams of the test cylinders. Fracture was initiated by means of hydraulic pressurization. One vessel was tested at each of three temperatures: 75, 130, and 190 0 F. Pretest analyses were made to predict the failure pressures and strains. Fracture toughness data obtained by equivalent-energy analysis of precracked Charpy-V tests and compact-tension specimen tests were used in the fracture analyses. The vessels behaved generally as had been expected. Posttest fracture analyses were also performed for each vessel. Detailed discussions of the fracture analysis methods developed in support of the vessel tests described are included. 34 references

  8. 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)

  9. 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)

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

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

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

  11. Burst pressure investigation of filament wound type IV composite pressure vessel

    Science.gov (United States)

    Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

    2017-12-01

    Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

  12. 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

  13. 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

  14. 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.

  15. Experience in surveillance of the prestress of concrete reactor vessels in Wylfa nuclear power station

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Walsh, S.R.

    1989-01-01

    This paper describes experience gained in the in-service surveillance of the prestressing system for the prestressed concrete reactor vessels (PCRVs) at Wylfa nuclear power station. The paper gives details of results for the prestressing system obtained from the statutory in-service inspection program of the PCRVs. The program includes a detailed examination of a selection of prestressing tendon anchorages, anchorage load checks using a lift-off technique on a one percent sample of tendons and corrosion inspection of samples of prestressing strand and determination of their mechanical properties. The results obtained from the above in-service inspections have shown that the prestressing system continues to function within its design limits

  16. Parametric Study on Important Variables of Aircraft Impact to Prestressed Concrete Containment Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sangshup; Hahm, Daegi; Choi, Inkil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    In this paper, to find the damage parameter, it is necessary to use many analysis cases and the time reduction. Thus, this paper uses a revised version of Riera's method. Using this method, the response has been found a Prestressed Concrete Containments Vessels (PCCVs) subject to impact loading, and the results of the velocity and mass of the important parameters have been analyzed. To find the response of the PCCVs subjected to aircraft impact load, it is made that a variable forcing functions depending on the velocity and fuel in the paper. The velocity variation affects more than fuel percentage, and we expect that the severe damage of the PCCVs with the same material properties is subject to aircraft impact load (more than 200m/s and 70%)

  17. 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)

  18. 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%).

  19. 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)

  20. 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)

  1. High stress monitoring of prestressing tendons in nuclear concrete vessels using fibre-optic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Perry, M., E-mail: marcus.perry@strath.ac.uk [Institute for Energy and Environment, University of Strathclyde, 204 George Street, Glasgow G1 1XW (United Kingdom); Yan, Z.; Sun, Z.; Zhang, L. [Aston Institute of Photonic Technologies, Aston University, Birmingham B4 7ET (United Kingdom); Niewczas, P. [Institute for Energy and Environment, University of Strathclyde, 204 George Street, Glasgow G1 1XW (United Kingdom); Johnston, M. [Civil Design Group, EDF Energy, Nuclear Generation, East Kilbride G74 5PG (United Kingdom)

    2014-03-15

    Highlights: • We weld radiation-resistant optical fibre strain sensors to steel prestressing tendons. • We prove the sensors can survive 1300 MPa stress (80% of steel's tensile strength). • Mechanical relaxation of sensors is characterised under 1300 MPa stress over 10 h. • Strain transfer between tendon and sensor remains at 69% after relaxation. • Sensors can withstand and measure deflection of tendon around a 4.5 m bend radius. - Abstract: Maintaining the structural health of prestressed concrete nuclear containments is a key element in ensuring nuclear reactors are capable of meeting their safety requirements. This paper discusses the attachment, fabrication and characterisation of optical fibre strain sensors suitable for the prestress monitoring of irradiated steel prestressing tendons. The all-metal fabrication and welding process allowed the instrumented strand to simultaneously monitor and apply stresses up to 1300 MPa (80% of steel's ultimate tensile strength). There were no adverse effects to the strand's mechanical properties or integrity. After sensor relaxation through cyclic stress treatment, strain transfer between the optical fibre sensors and the strand remained at 69%. The fibre strain sensors could also withstand the non-axial forces induced as the strand was deflected around a 4.5 m bend radius. Further development of this technology has the potential to augment current prestress monitoring practices, allowing distributed measurements of short- and long-term prestress losses in nuclear prestressed-concrete vessels.

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

    Science.gov (United States)

    Greulich, Owen Rudolf

    2017-01-01

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

  3. 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

  4. 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.

  5. 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)

  6. 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

  7. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    Science.gov (United States)

    Ellis, Tayera

    2015-01-01

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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. 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...

  19. 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...

  20. Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

    International Nuclear Information System (INIS)

    Li, D D; Jiang, J; Zhao, Z; Yi, W S; Lan, G

    2013-01-01

    We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system

  1. Modelling of pressure increase protection system for the vacuum vessel of W7-X device

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas, E-mail: tadas.kaliatka@lei.lt; Uspuras, Eugenijus; Kaliatka, Algirdas

    2016-11-01

    Highlights: • Two in-vessel LOCAs (partial and guillotine break of 40 mm diameter pipe of cooling system) for Wendelstein 7-X fusion device were analyzed. • The analysis of the processes in the cooling system, vacuum vessel and pressure increase protection system were performed using thermal-hydraulic RELAP5 Mod3.3 code. • The suitability of pressure increase protection system was assessed. - Abstract: In fusion devices, plasma is contained in a vacuum vessel. The vacuum vessel cannot withstand a pressure above atmospheric. Any damage of in-vessel components could lead to water ingress and may lead to pressure increase and possible damage of vacuum vessel. In order to avoid such undesirable consequences, the pressure increase protection system is designed. In this article, the processes occurring in the vacuum vessel and pressure increase protection system of W7-X device during LOCA (small and guillotine pipe break) event are analyzed. The model of W7-X cooling system, vacuum vessel and pressure increase protection system was developed using RELAP5 code. Numerical analysis of partial and guillotine break of 40 mm diameter pipe of cooling system was performed. Calculation results showed that burst disc of the pressure increase protection system does not open when the cross section area of partial break in the cooling system is smaller than 1 mm{sup 2}. During the guillotine break of cooling system, the burst disc opens, but pressure increase protection system is capable to prevent overpressure of the vacuum vessel.

  2. 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,...

  3. Development and application of a material law for steel-fibre-reinforced concrete with regard to its use for pre-stressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.; Borgerhoff, M.

    1995-01-01

    On the basis of the evaluation of many publications on the mechanical behaviour of steel fibre reinforced concrete (SFRC) and on the results of experiments using an SFRC especially developed for pre-stressed concrete reactor vessels (PCRVs), a material law for SFRC including general multiaxial stress conditions has been developed. From fibre pull-out tests described in the literature and by use of the experimental results, relations describing the capable tensile stress in SFRC after cracking, as a function of crack width, have been derived. There is a significant increase in the biaxial compressive strength of SFRC compared with plain concrete. The improved behaviour under multiaxial stress conditions, with one of the principal stresses being tensile, is outlined in comparison with different formulations of failure envelopes of plain concrete. For the purpose of verifying the material law implemented in the computer program used, analyses have been carried out for experiments with SFRC beams. After some modification concerning the shear behaviour, load-displacement curves and realistic crack propagations which correspond well have been obtained. In the stand-tube area in the centre of a PCRV top cap the use of SFRC is advantageous because of the difficulties concerning the arrangement of reinforcement in the concrete between the tubes. (orig.)

  4. 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.

  5. 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)

  6. 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)

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

    Science.gov (United States)

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

    2010-12-01

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

  8. 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

  9. 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

  10. Containment vs confinement trade study, small HTGR plant PCRV [prestressed concrete reactor vessel] concept

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-03-01

    This trade study has been conducted to evaluate the differences between four different HTGR nuclear power plants. All of the plants use a prestressed concrete reactor vessel (PCRV) to house the core and steam generation equipment. The reactor uses LEU U/Th fuel in prismatic carbon blocks. All plant concepts meet the utility/user requirements established for small HTGR plants. All plants will be evaluated with regard to their ability to produce safe, economical power to satisfy Goals 1, 2, and 3 of the HTGR program and by meeting the MUST criteria established in the concept evaluation plan. Capital costs for each plant were evaluated on a differential cost basis. These costs were developed according to the ``NUS`` code of accounts as defined in the Cost Estimating and Control Procedure, HP-20901. Accounts that were identical in scope for all four plants were not used for the comparison. Table 1-1 is a list of capital cost accounts that were evaluated for each plant.

  11. Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

    1978-02-01

    HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7.

  12. Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading

    International Nuclear Information System (INIS)

    Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

    1978-01-01

    HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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.

  18. 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

  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. Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

    OpenAIRE

    Young-Sun Choun; Junhee Park

    2015-01-01

    Background: Fibers have been used in cement mixture to improve its toughness, ductility, and tensile strength, and to enhance the cracking and deformation characteristics of concrete structural members. The addition of fibers into conventional reinforced concrete can enhance the structural and functional performances of safety-related concrete structures in nuclear power plants. Methods: The effects of steel and polyamide fibers on the shear resisting capacity of a prestressed concrete con...

  1. 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

  2. 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.)

  3. 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

  4. 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

  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.

    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.)

  6. 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.)

  7. 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)

  8. 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

  9. 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)

  10. 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.

  11. 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)

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

    Directory of Open Access Journals (Sweden)

    Hereil Pierre-Louis

    2015-01-01

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

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

    Science.gov (United States)

    Han, Hai-Chao

    2011-03-01

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

  14. 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

  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. Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

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

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

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

  19. 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

  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. On the Adequacy of API 521 Relief-Valve Sizing Method for Gas-Filled Pressure Vessels Exposed to Fire

    DEFF Research Database (Denmark)

    Andreasen, Anders; Nieto, Marcos Zan; Borroni, Filippo

    2018-01-01

    sense of security. Often the vessel wall will be weakened by high temperatures, before the PRV relieving pressure is reached. In this article, a multiparameter study has been performed taking into consideration various vessel sizes, design pressures (implicitly vessel wall thickness), vessel operating...

  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. Lay-out and construction of a pressure vessel built-up of cast steel segments for a pebble-bed high temperature reactor with a thermal power of 3000 MW

    International Nuclear Information System (INIS)

    Voigt, J.

    1978-03-01

    The prestressed cast vessel is an alternative to the prestressed concrete vessel for big high temperature reactors. In this report different cast steel vessel concepts for an HTR for generation of current with 3000 MW(th) are compared concerning their realization and economy. The most favourable variant serves as a base for the lay-out of the single vessel components as cast steel segments, bracing, cooling and outer sealing. Hereby the actual available possibilities of production and transport are considered. For the concept worked out possibilities of inspection and repair are suggested. A comparison of costs with adequate proposititons of the industry for a prestressed concrete and a cast iron pressure vessel investigates the economical competition. (orig.) [de

  4. 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)

  5. 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.

  6. Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

    International Nuclear Information System (INIS)

    Ventura, Mirta A.; Rosso, Ricardo D.

    2002-01-01

    In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

  7. An integrity evaluation method of the pressure vessel of nuclear reactors under pressurized thermal shock

    International Nuclear Information System (INIS)

    Matsubara, Masaaki; Okamura, Hiroyuki.

    1987-01-01

    Present paper proposes a new algorithm of the integrity evaluation of the pressure vessel of nuclear reactors under pressurized thermal shock, PTS. This method enables us to do an effective evaluation by superimposing proposed ''PTS state-transient curves'' and ''toughness transient curves'', and is superior to a conventional one in the following points; (1) easy to get an overall view of the result of PTS event for the variations of several parameters, (2) possible to evaluate a safety margin for irradiation embrittlement, and (3) enable to construct an Expert-friendly evaluation system. In addition, the paper shows that we can execute a safety assurance test by using a flat plate model with the same thickness as that of real plant. (author)

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

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

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

  9. Investigation of radial shear in the wall-base juncture of a 1:4 scale prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Dameron, R.A.; Rashid, Y.R.; Luk, V.K.; Hessheimer, M.F.

    1998-04-01

    Construction of a prestressed concrete containment vessel (PCCV) model is underway as part of a cooperative containment research program at Sandia National Laboratories. The work is co-sponsored by the Nuclear Power Engineering Corporation (NUPEC) of Japan and US Nuclear Regulatory Commission (NRC). Preliminary analyses of the Sandia 1:4 Scale PCCV Model have determined axisymmetric global behavior and have estimated the potential for failure in several areas, including the wall-base juncture and near penetrations. Though the liner tearing failure mode has been emphasized, the assumption of a liner tearing failure mode is largely based on experience with reinforced concrete containments. For the PCCV, the potential for shear failure at or near the liner tearing pressure may be considerable and requires detailed investigation. This paper examines the behavior of the PCCV in the region most susceptible to a radial shear failure, the wall-basemat juncture region. Prediction of shear failure in concrete structures is a difficult goal, both experimentally and analytically. As a structure begins to deform under an applied system of forces that produce shear, other deformation modes such as bending and tension/compression begin to influence the response. Analytically, difficulties lie in characterizing the decrease in shear stiffness and shear stress and in predicting the associated transfer of stress to reinforcement as cracks become wider and more extensive. This paper examines existing methods for representing concrete shear response and existing criteria for predicting shear failure, and it discusses application of these methods and criteria to the study of the 1:4 scale PCCV

  10. Investigation of radial shear in the wall-base juncture of a 1:4 scale prestressed concrete containment vessel model

    Energy Technology Data Exchange (ETDEWEB)

    Dameron, R.A.; Rashid, Y.R. [ANATECH Corp., San Diego, CA (United States); Luk, V.K.; Hessheimer, M.F. [Sandia National Labs., Albuquerque, NM (United States)

    1998-04-01

    Construction of a prestressed concrete containment vessel (PCCV) model is underway as part of a cooperative containment research program at Sandia National Laboratories. The work is co-sponsored by the Nuclear Power Engineering Corporation (NUPEC) of Japan and US Nuclear Regulatory Commission (NRC). Preliminary analyses of the Sandia 1:4 Scale PCCV Model have determined axisymmetric global behavior and have estimated the potential for failure in several areas, including the wall-base juncture and near penetrations. Though the liner tearing failure mode has been emphasized, the assumption of a liner tearing failure mode is largely based on experience with reinforced concrete containments. For the PCCV, the potential for shear failure at or near the liner tearing pressure may be considerable and requires detailed investigation. This paper examines the behavior of the PCCV in the region most susceptible to a radial shear failure, the wall-basemat juncture region. Prediction of shear failure in concrete structures is a difficult goal, both experimentally and analytically. As a structure begins to deform under an applied system of forces that produce shear, other deformation modes such as bending and tension/compression begin to influence the response. Analytically, difficulties lie in characterizing the decrease in shear stiffness and shear stress and in predicting the associated transfer of stress to reinforcement as cracks become wider and more extensive. This paper examines existing methods for representing concrete shear response and existing criteria for predicting shear failure, and it discusses application of these methods and criteria to the study of the 1:4 scale PCCV.

  11. 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)

  12. 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.

  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. 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

  15. 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%

  16. Blood pressure gradients in cerebral arteries: a clue to pathogenesis of cerebral small vessel disease.

    Science.gov (United States)

    Blanco, Pablo J; Müller, Lucas O; Spence, J David

    2017-09-01

    The role of hypertension in cerebral small vessel disease is poorly understood. At the base of the brain (the 'vascular centrencephalon'), short straight arteries transmit blood pressure directly to small resistance vessels; the cerebral convexity is supplied by long arteries with many branches, resulting in a drop in blood pressure. Hypertensive small vessel disease (lipohyalinosis) causes the classically described lacunar infarctions at the base of the brain; however, periventricular white matter intensities (WMIs) seen on MRI and WMI in subcortical areas over the convexity, which are often also called 'lacunes', probably have different aetiologies. We studied pressure gradients from proximal to distal regions of the cerebral vasculature by mathematical modelling. Blood flow/pressure equations were solved in an Anatomically Detailed Arterial Network (ADAN) model, considering a normotensive and a hypertensive case. Model parameters were suitably modified to account for structural changes in arterial vessels in the hypertensive scenario. Computations predict a marked drop in blood pressure from large and medium-sized cerebral vessels to cerebral peripheral beds. When blood pressure in the brachial artery is 192/113 mm Hg, the pressure in the small arterioles of the posterior parietal artery bed would be only 117/68 mm Hg. In the normotensive case, with blood pressure in the brachial artery of 117/75 mm Hg, the pressure in small parietal arterioles would be only 59/38 mm Hg. These findings have important implications for understanding small vessel disease. The marked pressure gradient across cerebral arteries should be taken into account when evaluating the pathogenesis of small WMIs on MRI. Hypertensive small vessel disease, affecting the arterioles at the base of the brain should be distinguished from small vessel disease in subcortical regions of the convexity and venous disease in the periventricular white matter.

  17. 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.

  18. Viscoelastic and thermal behavior of structural concrete with reference to containment vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1981-01-01

    A method of numerical viscoelastic stress analysis is described suitable for concrete structures operating at elevated temperatures. The paper describes how approximate numerical methods of elastic analysis of the finite element type can be extended to incorporate the viscoelastic behavior of structural concrete of the quasi-static type. A new eight parameter viscoelastic model is proposed to represent concrete behavior in the loaded and unloaded stage. The deformational expressions for the proposed viscoelastic analogue are also developed. Finally, as a result of courve-fitting procedures, the evaluation of the creep law coefficients are obtained for creep laws appropriate to a test regime. The proposed method is of general application providing that the properties of concrete are assessed reasonably well. The analytical predictions are compared with experimental results obtained on concrete model specimens loaded for 3 1/2 months, at a temperature of 80 0 C. (author)

  19. 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

  20. High Pressure Composite Overwrapped Pressure Vessel (COPV) Development Tests at Cryogenic Temperatures

    Science.gov (United States)

    Ray, David M.; Greene, Nathanael J.; Revilock, Duane; Sneddon, Kirk; Anselmo, Estelle

    2008-01-01

    Development tests were conducted to evaluate the performance of 2 COPV designs at cryogenic temperatures. This allows for risk reductions for critical components for a Gaseous Helium (GHe) Pressurization Subsystem for an Advanced Propulsion System (APS) which is being proposed for NASA s Constellation project and future exploration missions. It is considered an advanced system since it uses Liquid Methane (LCH4) as the fuel and Liquid Oxygen (LO2) as the oxidizer for the propellant combination mixture. To avoid heating of the propellants to prevent boil-off, the GHe will be stored at subcooled temperatures equivalent to the LO2 temperature. Another advantage of storing GHe at cryogenic temperatures is that more mass of the pressurized GHe can be charged in to a vessel with a smaller volume, hence a smaller COPV, and this creates a significant weight savings versus gases at ambient temperatures. The major challenge of this test plan is to verify that a COPV can safely be used for spacecraft applications to store GHe at a Maximum Operating Pressure (MOP) of 4,500 psig at 140R to 160R (-320 F to -300 F). The COPVs for these tests were provided by ARDE , Inc. who developed a resin system to use at cryogenic conditions and has the capabilities to perform high pressure testing with LN2.

  1. 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

  2. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    Science.gov (United States)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

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

  3. 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

  4. 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.

  5. 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)

  6. 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.)

  7. 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

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

    OpenAIRE

    Guo Han-yu; Tao Gang; Zhang Li-jing

    2016-01-01

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

  9. 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

  10. 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

  11. 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

  12. 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

  13. Positron affinity for precipitates in reactor pressure vessel steels

    Czech Academy of Sciences Publication Activity Database

    Brauer, G.; Puska, M. J.; Šob, Mojmír; Korhonen, T.

    1995-01-01

    Roč. 158, - (1995), s. 149-156 ISSN 0029-5493. [Meeting of the International Group on Radiation Damage Mechanisms in Preassure Vessel Steels /5./. Santa Barbara, 02.05.1994-06.05.1994] Impact factor: 0.134, year: 1995

  14. Comparison of closed-pressurized and open-refluxed vessel ...

    African Journals Online (AJOL)

    Samples of residual fuel oil reference material (SRM 1634c) were mineralized in closed digestion vessels from Milestone Laboratory Systems (MLS) or from PAAR (HPA) or in open-refluxed microwave digestion flasks from Prolabo. The three digestion systems were evaluated in terms of accuracy and precision, reagents ...

  15. AWWA C303-17 concrete pressure pipe, bar-wrapped, steel-cylinder type

    CERN Document Server

    2017-01-01

    This standard describes the manufacture of concrete pressure pipe, reinforced with a steel cylinder that is helically wrapped with mild steel bar reinforcement, in sizes ranging from 10 in. through 72 in. (250 mm through 1,830 mm), inclusive, and for working pressures up to 400 psi (2,760 kPa).

  16. 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)

  17. 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.

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

    Science.gov (United States)

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

    2017-01-01

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

  19. Water Permeability of Pervious Concrete Is Dependent on the Applied Pressure and Testing Methods

    Directory of Open Access Journals (Sweden)

    Yinghong Qin

    2015-01-01

    Full Text Available Falling head method (FHM and constant head method (CHM are, respectively, used to test the water permeability of permeable concrete, using different water heads on the testing samples. The results indicate the apparent permeability of pervious concrete decreasing with the applied water head. The results also demonstrate the permeability measured from the FHM is lower than that from the CHM. The fundamental difference between the CHM and FHM is examined from the theory of fluid flowing through porous media. The testing results suggest that the water permeability of permeable concrete should be reported with the applied pressure and the associated testing method.

  20. 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.

  1. 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)

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

    International Nuclear Information System (INIS)

    O'Neil, R.

    1977-01-01

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

  3. 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

  4. Pressure controlled clamp using shape memory alloy for minimal vessel invasion in blood flow occlusion.

    Science.gov (United States)

    Zhang, Ye; Kanetaka, Hiroyasu; Sano, Yuya; Kano, Mitsuhiro; Kudo, Tadaaki; Sato, Takumi; Shimizu, Yoshinaka

    2013-01-01

    Vessel damage after clamping may affect the success of surgical operations. A new pressure controlled clamp (SMA clamp) was designed using super elastic property of shape memory alloy (SMA) to realize atraumatic vessel occlusion. The ability and biological effect of the SMA clamp to control pressure was investigated in vivo. The loading-displacement curves of the SMA clamps (experimental group) and conventional clamp (control group) by occlusion of pig carotid arteries were evaluated using a clamping-pressure analyzing system. To investigate macroscopically and histologically the vessel damage of the SMA and conventional clamps, pig carotid arteries were stained with Evan's blue and its histological sections were stained with Elastica Massion after clamping for fifteen minutes. Constant value was shown in the loading-displacement curve of SMA clamp. In the control group, damaged area stained with Evan's blue in the vessel wall showed enlargement with the pressure increasing. Less areas in experimental groups are observed than that in the control group. Histological section in the experimental group showed no obvious except a slight compressive damage in the tunica intima. In the control group, vessel wall showed irreversible damages. This experiment indicated that the SMA clamp, which has a unique mechanical property, can be used without vessels damage. This pressure controlled clamp can be a selection in clinical apparatus to improve surgical safety.

  5. Non-Axisymmetric Inflatable Pressure Structure (NAIPS) Concept that Enables Mass Efficient Packageable Pressure Vessels with Sealable Openings

    Science.gov (United States)

    Doggett, William R.; Jones, Thomas C.; Kenner, Winfred S.; Moore, David F.; Watson, Judith J.; Warren, Jerry E.; Makino, Alberto; Yount, Bryan; Selig, Molly; Shariff, Khadijah; hide

    2016-01-01

    Achieving minimal launch volume and mass are always important for space missions, especially for deep space manned missions where the costs required to transport mass to the destination are high and volume in the payload shroud is limited. Pressure vessels are used for many purposes in space missions including habitats, airlocks, and tank farms for fuel or processed resources. A lucrative approach to minimize launch volume is to construct the pressure vessels from soft goods so that they can be compactly packaged for launch and then inflated en route or at the final destination. In addition, there is the potential to reduce system mass because the packaged pressure vessels are inherently robust to launch loads and do not need to be modified from their in-service configuration to survive the launch environment. A novel concept is presented herein, in which sealable openings or hatches into the pressure vessels can also be fabricated from soft goods. To accomplish this, the structural shape is designed to have large regions where one principal stress is near zero. The pressure vessel is also required to have an elongated geometry for applications such as airlocks.

  6. Study of the concrete tensile creep: application for the containment vessel of the nuclear power plants (PWR)

    International Nuclear Information System (INIS)

    Reviron, Nanthilde

    2009-01-01

    The aim of this work is to study experimentally and to conduct numerical simulations on the creep of concrete subjected to tensile stresses. The main purpose is to predict the behaviour of containment vessels of nuclear power plants (PWR) in the case of decennial test or accident. In order to satisfy to these industrial needs, it is necessary to characterize the behaviour of concrete under uniaxial tension. Thus, an important experimental study of tensile creep in concrete has been performed for different loading levels (50%, 70% and 90% of the tensile strength). In these tests, load was kept constant during 3 days. Several tests were performed: measurements of elastic properties and strength (in tension and in compression), monitoring of drying, shrinkage, basic creep and drying creep strains. Moreover, compressive creep tests were also performed and showed a difference with tensile creep. Furthermore, decrease of tensile strength and failure under tensile creep for large loading levels were observed. A numerical model has been proposed and developed in Cast3m finite element code. (author)

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-09-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-15

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

  13. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  14. 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

  15. 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

  16. 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...

  17. 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)

  18. 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

  19. Application of Non-pressure Reinforced Concrete Pipes in Modern Construction and Reconstruction of Highways

    Science.gov (United States)

    Rakitin, B. A.; Pogorelov, S. N.; Kolmogorova, A. O.

    2017-11-01

    Modern highway construction technologies provide for the quality water discharge systems to increase facilities’ service life. Pipeline operating conditions require the use of durable and reliable materials and structures. The experience in using reinforced concrete pipes for these purposes shows their utilization efficiency. The present paper considers the experience in the use of non-pressure reinforced concrete pipes manufactured by the German company SCHLOSSER-PFEIFFER under the Ural region geological and climatic conditions. The authors analyzed the actual operation of underground pipelines and effective loads upon them. A detailed study of the mechanical properties of reinforced concrete pipes is necessary to improve their production technology and to enhance their serviceability. The use of software-based methods helped to develop a mathematical model and to estimate the strength and crack resistance of reinforced concrete pipes at different laying depths. The authors carried out their complex research of the strain-stress behaviour of reinforced concrete pipes and identified the most hazardous sections in the structure. The calculations performed were confirmed by the results of laboratory tests completed in the construction materials, goods, and structures test center. Based on the completed research, the authors formulated their recommendations to improve the design and technology of non-pressure reinforced concrete pipes.

  20. 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

  1. 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

  2. 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

  3. 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

  4. Effects of dynamic shear and transmural pressure on wall shear stress sensitivity in collecting lymphatic vessels

    Science.gov (United States)

    Kornuta, Jeffrey A.; Nepiyushchikh, Zhanna; Gasheva, Olga Y.; Mukherjee, Anish; Zawieja, David C.

    2015-01-01

    Given the known mechanosensitivity of the lymphatic vasculature, we sought to investigate the effects of dynamic wall shear stress (WSS) on collecting lymphatic vessels while controlling for transmural pressure. Using a previously developed ex vivo lymphatic perfusion system (ELPS) capable of independently controlling both transaxial pressure gradient and average transmural pressure on an isolated lymphatic vessel, we imposed a multitude of flow conditions on rat thoracic ducts, while controlling for transmural pressure and measuring diameter changes. By gradually increasing the imposed flow through a vessel, we determined the WSS at which the vessel first shows sign of contraction inhibition, defining this point as the shear stress sensitivity of the vessel. The shear stress threshold that triggered a contractile response was significantly greater at a transmural pressure of 5 cmH2O (0.97 dyne/cm2) than at 3 cmH2O (0.64 dyne/cm2). While contraction frequency was reduced when a steady WSS was applied, this inhibition was reversed when the applied WSS oscillated, even though the mean wall shear stresses between the conditions were not significantly different. When the applied oscillatory WSS was large enough, flow itself synchronized the lymphatic contractions to the exact frequency of the applied waveform. Both transmural pressure and the rate of change of WSS have significant impacts on the contractile response of lymphatic vessels to flow. Specifically, time-varying shear stress can alter the inhibition of phasic contraction frequency and even coordinate contractions, providing evidence that dynamic shear could play an important role in the contractile function of collecting lymphatic vessels. PMID:26333787

  5. Evaluation of seismic shear capacity of prestressed concrete containment vessels with fiber reinforcement

    Directory of Open Access Journals (Sweden)

    Young-Sun Choun

    2015-10-01

    Conclusion: The addition of fibers into conventional concrete increases the ductility and energy dissipation of wall structures significantly. Fibers can be effectively used to improve the structural performance of a PCCV subjected to strong ground motions. Steel fibers are more effective in enhancing the shear performance of a PCCV than polyamide fibers.

  6. A novel high pressure, high temperature vessel used to conduct long-term stability measurements of silicon MEMS pressure transducers

    Science.gov (United States)

    Wisniewiski, David

    2014-03-01

    The need to quantify and to improve long-term stability of pressure transducers is a persistent requirement from the aerospace sector. Specifically, the incorporation of real-time pressure monitoring in aircraft landing gear, as exemplified in Tire Pressure Monitoring Systems (TPMS), has placed greater demand on the pressure transducer for improved performance and increased reliability which is manifested in low lifecycle cost and minimal maintenance downtime through fuel savings and increased life of the tire. Piezoresistive (PR) silicon MEMS pressure transducers are the primary choice as a transduction method for this measurement owing to their ability to be designed for the harsh environment seen in aircraft landing gear. However, these pressure transducers are only as valuable as the long-term stability they possess to ensure reliable, real-time monitoring over tens of years. The "heart" of the pressure transducer is the silicon MEMS element, and it is at this basic level where the long-term stability is established and needs to be quantified. A novel High Pressure, High Temperature (HPHT) vessel has been designed and constructed to facilitate this critical measurement of the silicon MEMS element directly through a process of mechanically "floating" the silicon MEMS element while being subjected to the extreme environments of pressure and temperature, simultaneously. Furthermore, the HPHT vessel is scalable to permit up to fifty specimens to be tested at one time to provide a statistically significant data population on which to draw reasonable conclusions on long-term stability. With the knowledge gained on the silicon MEMS element, higher level assembly to the pressure transducer envelope package can also be quantified as to the build-effects contribution to long-term stability in the same HPHT vessel due to its accommodating size. Accordingly, a HPHT vessel offering multiple levels of configurability and robustness in data measurement is presented, along

  7. Calculation method for residual stress analysis of filament-wound spherical pressure vessels

    International Nuclear Information System (INIS)

    Knight, C.E. Jr.

    1976-01-01

    Filament wound spherical pressure vessels may be produced with very high performance factors. These performance factors are a calculation of contained pressure times enclosed volume divided by structure weight. A number of parameters are important in determining the level of performance achieved. One of these is the residual stress state in the fabricated unit. A significant level of an unfavorable residual stress state could seriously impair the performance of the vessel. Residual stresses are of more concern for vessels with relatively thick walls and/or vessels constructed with the highly anisotropic graphite or aramid fibers. A method is established for measuring these stresses. A theoretical model of the composite structure is required. Data collection procedures and techniques are developed. The data are reduced by means of the model and result in the residual stress analysis. The analysis method can be used in process parameter studies to establish the best fabrication procedures

  8. 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.)

  9. 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.)

  10. 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.)

  11. 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.)

  12. 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

  13. 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.)

  14. 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)

  15. Acoustic emission test on a 25mm thick mild steel pressure vessel with inserted defects

    International Nuclear Information System (INIS)

    Bentley, P.G.; Dawson, D.G.; Hanley, D.J.; Kirby, N.

    1976-12-01

    Acoustic emission measurements have been taken on an experimental mild steel vessel with 4 inserted defects ranging in severity up to 90% of through thickness. The vessel was subjected to a series of pressure excursions of increasing magnitude until failure occurred by extension of the largest inserted defect through the vessel wall. No acoustic emission was detected throughout any part of the tests which would indicate the presence of such serious defects or of impending failure. Measurements of acoustic emission from metallurgical specimens are included and the results of post test inspection using conventional NDT and metallographic techniques are reported. (author)

  16. 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)

  17. Managing Pressure Vessel Equipment as a Capital Asset.

    Science.gov (United States)

    Robinson, Glenn; Trombley, Robert; Shultes, Kenneth

    1999-01-01

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

  18. 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...

  19. 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

  20. 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

  1. 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

  2. Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events

    International Nuclear Information System (INIS)

    Hirano, Masashi; Kohsaka, Atsuo.

    1992-02-01

    A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k I ) with the crack initiation toughness (k Ic ) and crack arrest toughness (k Ic ), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author)

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. Applicability of newly developed 610MPa class heavy thickness high strength steel to boiler pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Katayama, Norihiko; Kaihara, Shoichiro; Ishii, Jun [Ishikawajima-Harima Heavy Industries Corp., Yokohama (Japan); Kajigaya, Ichiro [Ishikawajima-Harima Heavy Industries Corp., Tokyo (Japan); Totsuka, Takehiro; Miyazaki, Takashi [Ishikawajima-Harima Heavy Industries Corp., Aioi (Japan)

    1995-11-01

    Construction of a 350 MW Class PFBC (Pressurized Fluidized Bed Combustion) boiler plant is under planning in Japan. Design temperature and pressure of the vessel are maximum 350 C and 1.69 MPa, respectively. As the plate thickness of the vessel exceeds over 100 mm, high strength steel plate of good weldability and less susceptible to reheat cracking was required and developed. The steel was aimed to satisfy the tensile strength over 610 MPa at 350 C after postweld heat treatment (PWHT), with good notch toughness. The authors investigated the welding performances of the newly developed steel by using 150 mm-thick plate welded by pulsed-MAG and SAW methods. It was confirmed that the newly developed steel and its welds possess sufficient strength and toughness after PWHT, and applicable to the actual pressure vessel.

  10. 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

  11. Study on effective prestressing effects on concrete containment under the design-basis pressure condition

    International Nuclear Information System (INIS)

    Sun Feng; Pan Rong; Wang Lu; Mao Huan; Yang Yu

    2013-01-01

    Prestressing technology is widely used in nuclear power plant containment building, and the durability of containment structure is affected directly by the distribution and loss of prestressing value under design-basis pressure. Containment structure and the distribution of prestressing system are introduced briefly. Furthermore, the calculating process of horizontal prestressing bunch loss near the equipment hatch hole is put forward in details, and the containment structure prestressing loss when 5-year pressure test is obtained. Based above analysis, the finite element model of the prestressed concrete containment structure is built by using ANSYS code, the prestressing effect on concrete containment is analysed. The results show that most of the design pressure is bore by the prestressing system under the design-basis pressure, so the containment structure is safe. These conclusions are consistent with prestressing containment system design concepts, which can provide reference to the engineering staff. (authors)

  12. Dynamic Increase Factors for High Performance Concrete in Compression using Split Hopkinson Pressure Bar

    DEFF Research Database (Denmark)

    Riisgaard, Benjamin; Ngo, Tuan; Mendis, Priyan

    2007-01-01

    This paper provides dynamic increase factors (DIF) in compression for two different High Performance Concretes (HPC), 100 MPa and 160 MPa, respectively. In the experimental investigation 2 different Split Hopkinson Pressure Bars are used in order to test over a wide range of strain rates, 100 sec1...

  13. 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.

  14. Durability and safety of concrete structures in the nuclear context. The case of the containment vessel

    International Nuclear Information System (INIS)

    Torrenti, J.M.; Nahas, G.

    2011-01-01

    The durability of structures, because of its economic and environmental implications, is one of the actual hot topics in civil engineering. In the field of nuclear energy, we are facing very challenging problems like: how could we prolong the service life of actual nuclear containments and how can we assure the durability of a radioactive storage on the very long term (several centuries)? These already difficult questions in a classical civil engineering view are even more complicated in the field of nuclear energy where the structures are massive and the safety of the installations has to be considered. For the containment of nuclear power plants, these stakes will be lit with some examples of research concerning the mechanical behaviour of concrete and concrete structures (at early age, in service on long scales of time and in the event of an accident), the durability of the concrete structures (leaching, swelling due to delayed ettringite formation - DEF -) and the couplings between mechanics and durability. Finally, the importance of probabilistic aspects and the inherent difficulties will be shown. (authors)

  15. 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...

  16. 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...

  17. The design of lifting attachments for the erection of large diameter and heavy wall pressure vessels

    International Nuclear Information System (INIS)

    Antalffy, Leslie P.; Miller, George A.; Kirkpatrick, Kenneth D.; Rajguru, Anil; Zhu, Yong

    2016-01-01

    Lifting attachments for the erection of large diameter and heavy wall pressure vessels require special consideration to ensure that their attachment to their vessel shells or heads do not overstress the vessel during the erection process when lifting these from grade onto their respective foundations. Today, in refinery and petrochemical services, large diameter vessels with diameters ranging up to 15 m and reactors with lifting weights in the range of 700–1400 tons are not uncommon. In today's fabrication market, these vessels may be purchased and fabricated in shops dispersed globally and will require unique equipment for their safe handling, transportation and subsequent erection. The challenge is to design the lifting attachments in such a manner that the attachments provide a safe, cost effective and effective solution based upon the limitations of the job site lift equipment available for erection. Such equipment for the transportation and subsequent lifting of large diameter and heavy wall pressure equipment is usually scarce and quite expensive. Planning ahead, well in advance of the lift date is almost a mandatory requirement. Usually, the specific parameters of the vessel to be lifted and the lifting equipment available at the site will dictate the type of lifting attachments to be designed for the vessel. Once the type of vessel attachment has been chosen, careful consideration must be given to the design of attachments to the pressure vessel in consideration to ensure that the vessel and lifting components are not overstressed during the lifting process. The paper also discusses different types of lifting attachments that may be attached to each end of the vessel either by bolting or welding and discusses the pros and cons of each. The paper also provides an example of a finite element analysis (FEA) of a top nozzle, a FEA of a pair of lifting trunnions and a FEA of welded on lifting lugs for buried pipe. The purpose of the paper is to outline

  18. Control method for pool water of pressure suppression chamber in reactor container vessel

    International Nuclear Information System (INIS)

    Matsuzaki, Tetsuo; Konomaru, Toshimi; Saito, Koichi.

    1996-01-01

    In a reactor container having a pressure suppression chamber in adjacent with the outer circumference of a pedestal of a reactor pressure vessel, at least a portion of pool water is provisionally transported from the pressure suppression chamber to the inner side surrounded by the pedestal of the pressure vessel, and stored therein. In addition, an opening of the side wall of the pedestal is closed to raise the water level of the provisionally stored water thereby increasing the amount of provisionally stored pool water. Predetermined operations are performed in the pressure suppression chamber after transporting the pool water to the inner side surrounded by the pedestal of the pressure vessel. Namely, a portion of the pool water of the pressure suppression chamber is transported to the inner space surrounded by the pedestal to provisionally store it thereby forming a circumstance for enabling predetermined operations such as inspection and re-coating in the pressure suppression chamber. Then, radiation contamination prevailing to the outer side of the reactor container can be reduced thereby enabling to obtain various effects, namely, reduction in the amount of equipments for provisional installment, shortening for construction term and decrease in the amount of waste materials. (I.S.)

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

    Science.gov (United States)

    Russell, Rick; Skow, Miles

    2013-01-01

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

  20. 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

  1. 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

  2. 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.)

  3. 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.

  4. 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.

  5. 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...

  6. Mesos-scale modeling of irradiation in pressurized water reactor concrete biological shields

    Energy Technology Data Exchange (ETDEWEB)

    Le Pape, Yann [ORNL; Huang, Hai [Idaho National Laboratory (INL)

    2016-01-01

    Neutron irradiation exposure causes aggregate expansion, namely radiation-induced volumetric expansion (RIVE). The structural significance of RIVE on a portion of a prototypical pressurized water reactor (PWR) concrete biological shield (CBS) is investigated by using a meso- scale nonlinear concrete model with inputs from an irradiation transport code and a coupled moisture transport-heat transfer code. RIVE-induced severe cracking onset appears to be triggered by the ini- tial shrinkage-induced cracking and propagates to a depth of > 10 cm at extended operation of 80 years. Relaxation of the cement paste stresses results in delaying the crack propagation by about 10 years.

  7. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

  8. Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rossinski, S.T.; Carter, R.G.

    1996-07-01

    The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

  9. A quantitative methodology for reactor vessel pressurized thermal shock decision making

    International Nuclear Information System (INIS)

    Ackerson, D.S.; Balkey, K.R.; Meyer, T.A.; Ofstun, R.P.; Rupprecht, S.D.; Sharp, D.R.

    1983-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Considerations of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS. A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. (orig./RW)

  10. Analysis of pressurized resistance vessel diameter changes with a low cost digital image processing device.

    Science.gov (United States)

    Fischer, J G; Mewes, H; Hopp, H H; Schubert, R

    1996-06-01

    A low cost digital image processing device (frame grabber) together with a program running under MS_WINDOWS for automatic on-line analysis of diameter changes of in vitro pressurized blood vessels with an inner diameter of 80-400 microns is presented. The frame grabber is designed to receive light microscopic images either from a video camera or from a VCR and to present the digitized image on the computer monitor. The special software allows to manipulate the image, e.g. filtering, calibrating, storing of vessel images, and detects the outer and inner border of the two vessel walls with a new, simple algorithm. The inner diameter and the vessel wall thickness are calculated and the diameter is presented in a diameter versus time diagram on the monitor screen. Further, these data are stored in an ASCII-file for later import into calculation and presentation programs like MS-EXCEL.

  11. An analysis of reactor pit pressurization and forces applied on reactor vessel

    International Nuclear Information System (INIS)

    Wang Rongzhong; Li Feng

    1997-12-01

    The pressure and temperature transients with the time of the reactor pit during LOCA have been analyzed by using Catem computer code for Qinshan-2 nuclear power plant. The force and bending moment on the inlet and outlet nozzles of the reactor vessel also have been calculated by using Wformom code. Qinshan-2 NPP is a two-loop nuclear power plant. The cold water of the accumulators are directly injected into the downcomer of reactor vessel. Injection line of accumulators is located at the same level with the inlet and outlet nozzles. These geometry characteristics have been taken into account in the circumferential vessel pit nodding using five volumes around the vessel. The assumptions used in the analysis and calculation results have been presented. Many sensitive calculations have been performed for different break size and circumferential nodding

  12. Effects of low upper shelf fracture toughness on reactor vessel integrity during pressurized thermal shock events

    International Nuclear Information System (INIS)

    Bamford, W.H.; Heinecke, C.C.; Balkey, K.R.

    1988-01-01

    For the past decade, significant attention has been focused on the subject of nuclear rector vessel integrity during pressurized thermal shock (PTS) events. The issue of low upper shelf fracture toughness at operating temperatures has been a consideration for some reactor vessel materials since the early 1970's. Deterministic and probabilistic fracture mechanics sensitivity studies have been completed to evaluate the interaction between the PTS and lower upper shelf toughness issues that result from neutron embrittlement of the critical beltline region materials. This paper presents the results of these studies to show the interdependency of these fracture considerations in certain instances and to identify parameters that need to be carefully treated in reactor vessel integrity evaluations for these subjects. This issue is of great importance to those vessels which have low upper shelf toughness, both for demonstrating safety during the original design life and in life extension assessments

  13. Damage-tolerant design and inspection philosophy for nuclear and other pressure vessels

    International Nuclear Information System (INIS)

    Adams, N.J.I.

    1980-01-01

    Statistical analyses of pressure vessel failure rates indicate that, to date, the record is very good. However, the public hazard and environmental consequences of failure in certain industrial processes now give cause for much greater concern. With the exception of an Appendix in ASME III, the current design codes and requirements for new vessels are all based on the assumption that they are free from cracklike defects, but engineers recognize tht such perfect vessels cannot be manufactured. Taking into account failure mechanisms, material properties, pre- and in-service inspection, proof testing, failure statistics and probabilistic methods, views are put forward on how a damage-tolerant design and inspection philosophy may be developed to reduce further the possibility of ''rogue'' vessel failure. 21 refs

  14. Minimum critical crack depths in pressure vessels guidelines for nondestructive testing

    International Nuclear Information System (INIS)

    Crossley, M.R.; Townley, C.H.A.

    1983-09-01

    Estimates of the minimum critical depths which can be expected in high quality vessels designed to certain British and American Code rules are given. A simple means of allowing for fatigue crack growth in service is included. The data which are presented can be used to decide what sensitivity and what reporting levels should be employed during an ultrasonic inspection of a pressure vessel. It is emphasised that the minimum crack depths are those which would be relevant to a vessel in which the material is stressed to its maximum permitted value during operation. Stresses may, in practice, be significantly less than this. Less restrictive inspection standards may be established, if it were considered worthwhile to carry out a detailed stress analysis of the particular vessel under examination. (author)

  15. The probabilistic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China); Lu, Feng; Wang, Rongshan; Yu, Weiwei [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China); Wang, Donghui [State Nuclear Power Plant Service Company, 200237 Shanghai (China); Zhang, Guodong; Xue, Fei [Suzhou Nuclear Power Research Institute, 215004 Suzhou, Jiangsu Province (China)

    2015-12-01

    Highlights: • The methodology and the case study of the FAVOR software were shown. • The over-conservative parameters in the DFM were shown. • The differences between the PFM and the DFM were discussed. • The limits in the current FAVOR were studied. - Abstract: The pressurized thermal shock (PTS) event poses a potentially significant challenge to the structural integrity of the reactor pressure vessel (RPV) during the long time operation (LTO). In the USA, the “screening criteria” for maximum allowable embrittlement of RPV material, which forms part of the USA regulations, is based on the probabilistic fracture mechanics (PFM). The FAVOR software developed by Oak Ridge National Laboratory (ORNL) is used to establish the regulation. As the technical basis of FAVOR is not the most widely-used and codified methodologies, such as the ASME and RCC-M codes, in most countries (with exception of the USA), proving RPV integrity under the PTS load is still based on the deterministic fracture mechanics (DFM). As the maximum nil-ductility-transition temperature (RT{sub NDT}) of the beltline material for the 54 French RPVs after 40 years operation is higher than the critical values in the IAEA-TECDOC-1627 and European NEA/CSNI/R(99)3 reports (while still obviously lower than the “screening criteria” of the USA), it may conclude that the RPV will not be able to run in the LTO based on the DFM. In the FAVOR, the newest developments of fracture mechanics are applied, such as the warm pre-stress (WPS) effect, more accurate estimation of the flaw information and less conservation of the toughness (such as the three-parameter Weibull distribution of the fracture toughness). In this paper, the FAVOR software is first applied to show both the methodology and the results of the PFM, and then the limits in the current FAVOR software (Version 6.1, which represents the baseline for re-assessing the regulation of 10 CFR 50.61), lack of the impact of the constraint effect

  16. 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

  17. Development of Improved Composite Pressure Vessels for Hydrogen Storage

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-29

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

  18. A rigid disc for protection of exposed blood vessels during negative pressure wound therapy.

    Science.gov (United States)

    Anesäter, Erik; Borgquist, Ola; Torbrand, Christian; Roupé, K Markus; Ingemansson, Richard; Lindstedt, Sandra; Malmsjö, Malin

    2013-02-01

    There are increasing reports of serious complications and deaths associated with negative pressure wound therapy (NPWT). Bleeding may occur when NPWT is applied to a wound with exposed blood vessels. Inserting a rigid disc in the wound may protect these structures. The authors examined the effects of rigid discs on wound bed tissue pressure and blood flow through a large blood vessel in the wound bed during NPWT. Wounds were created over the femoral artery in the groin of 8 pigs. Rigid discs were inserted. Wound bed pressures and arterial blood flow were measured during NPWT. Pressure transduction to the wound bed was similar for control wounds and wounds with discs. Blood flow through the femoral artery decreased in control wounds. When a disc was inserted, the blood flow was restored. NPWT causes hypoperfusion in the wound bed tissue, presumably as a result of mechanical deformation. The insertion of a rigid barrier alleviates this effect and restores blood flow.

  19. A prediction method for long-term behavior of prestressed concrete containment vessels

    International Nuclear Information System (INIS)

    Ozaki, M.; Abe, T.; Watanabe, Y.; Kato, A.; Yamaguchi, T.; Yamamoto, M.

    1995-01-01

    This paper presents results of studies on the long-term behavior of PCCVs at Taruga Unit No 2 and Ohi Unit No 3/4 power stations. The objective of this study is to evaluate the measured strain in the concrete and reduction force in the tendons, and to establish the prediction methods for long-term PCCVs behavior. Comparing the measured strains with those calculated due to creep and shrinkage of the concrete, those in contrast were investigated. Furthermore, the reduced tendon forces are calculated considering losses in elasticity, relaxation, creep and shrinkage. The measured reduction in the tendon forces is compared with the calculated. Considering changes in temperature and humidity, the measured strains and tendon forces were in good agreement with those calculated. From the above results, it was confirmed that the residual pre stresses in the PCCVs maintain the predicted values at the design stage, and that the prediction method of long-term behaviors has sufficient reliability. (author). 10 refs., 8 figs., 3 tabs

  20. The use of high-hydrostatic pressure treatment to decellularize blood vessels.

    Science.gov (United States)

    Funamoto, Seiichi; Nam, Kwangwoo; Kimura, Tsuyoshi; Murakoshi, Ayako; Hashimoto, Yoshihide; Niwaya, Kazuo; Kitamura, Soichiro; Fujisato, Toshiya; Kishida, Akio

    2010-05-01

    A decellularization method using high-hydrostatic pressure (HHP) technology (>600MPa) is described. The HHP disrupts the cells inside the tissue. The cell debris can be eliminated with a simple washing process, producing clean, decellularized tissue. In this study, porcine aortic blood vessel was decellularized by HHP. The mechanical properties and in vivo performance of the decellularized tissue were evaluated. Mechanical properties of the decellularized tissue were not altered by the HHP treatment. Reduced inflammation of the decellularized tissue was confirmed by xenogenic transplant experimentation. An allogenic transplantation study showed that decellularized blood vessel endured the arterial blood pressure, and there was no clot formation on the luminal surface. In addition, cellular infiltration into the vessel wall was observed 4 weeks after implantation, suggesting that HHP treatments could be applied widely as a high-quality decellularization method. Copyright 2010 Elsevier Ltd. All rights reserved.