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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. Rapid construction of concrete pressure vessels

    International Nuclear Information System (INIS)

    Limbert, D.; Weatherseed, D.C.

    1989-01-01

    This paper opens with a general description of the concrete pressure vessel followed by a more detailed examination of the critical elements of the construction, including choice of methods and plant which were selected to ensure its rapid construction. The pressure vessel construction cannot be treated in isolation, because it is very closely linked with its surrounding structures - namely the reactor hall which surrounds it and the charge hall which tops it, as will be seen in the context of this paper. Rate of progress of construction is not entirely in the civil contractor's hands because so many of the operations affecting the civil works are of a mechanical nature, hence a very close liaison and understanding amongst all contractors concerned was of the utmost importance. (author)

  3. Method of detecting construction faults in concrete pressure vessels

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  4. Prestressed concrete pressure vessels for nuclear reactors - 1973

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

  5. 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

  6. 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.)

  7. 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)

  8. 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)

  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. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  11. 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)

  12. The evolution and structural design of prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Hannah, I.W.

    1978-01-01

    The introduction of the prestressed concrete pressure vessel to contain the main gas coolant circuit of nuclear reactors has marked a major step forward. This chapter traces the evolution and development of the PCPV, and lists the principal parameters adopted. Current design and loading standards are discussed in relation to the two main limit states of serviceability and safety. Prestressed concrete pressure vessel analysis has called for very extensive adaptation and expansion of conventional finite element and finite difference methods in order to deal with the elevated temperature of operation, together with extensive concrete testing at temperature and under multi-directional stressing. These new methods and extra data are being adopted in prestressed applications in other fields and may well prove to be of much wider significance than is presently appreciated. (author)

  13. 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)

  14. Heat and mass transfer in a concrete pressure vessel

    International Nuclear Information System (INIS)

    Zangle, K.; Sadouki, H.; Wittmann, F.H.

    1989-01-01

    Pressure vessels of prestressed concrete for high temperature reactors are subjected to high mechanical and thermal stresses during the reactors normal working conditions and in particular accidental conditions. According to a large temperature gradient between the inner liner and the outer side of the thickwalled vessel, physical as well as chemical processes take place in concrete. Temperature and moisture content of concrete have a big influence on these processes. During the last years different investigations have been conducted in order to determine characteristic values of concrete under these conditions. At present the authors conduct a series of experiments on model vessels of prestressed concrete and a large number of small specimens. The aims of these tests can be briefly summarized as follows: experimental determination of transport coefficients for a numerical analysis; determination of chemical reactions under hydrothermal conditions and their significance for the risk of corrosion; determination of temperature and moisture distribution as a function of time; and determination of the strength development in the zones subjected to elevated temperatures

  15. 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.)

  16. Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1981-01-01

    Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

  17. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

    The numerical procedures for predicting the nonlinear behavior 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 linear 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 elasti-plastic behavior of the liner and reinforcing steel. Various failure theories for concrete have been proposed recently. Also, there are alternative strategies for solving the discrete system equations over the design life, accounting for test loads, pressure and temperature operational loads, creep unloading and abnormal loads. The proposed methods are reviewed, and a new formulation developed by the authors is described. A number of comparisons with experimental tests results and other numerical schemes are presented. These examples demonstrate the validity of the formulation and also provide valuable information concerning the cost and accuracy of the various solution strategies i.e., total vs. incremental loading and initial vs. tangent stiffness. Finally, 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

  18. 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

  19. 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

  20. High temperature helium test rig with prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schmidl, H.

    1975-10-01

    The report gives a short description of the joint project prestressed concrete vessel-helium test station as there is the building up of the concrete structure, the system of instrumentation, the data processing, the development of the helium components as well as the testing programs. (author)

  1. 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

  2. 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.)

  3. Containers, particularly prestressed concrete pressure vessels for nuclear reactor plants

    International Nuclear Information System (INIS)

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

    1986-01-01

    Pressure and temperature changes act on the liner, which cause differential expansion between the liner and the prestressed concrete. So that there will be no overload or damage to the liner, its anchoring or the concrete structure, cutouts are provided in the concrete at deflection positions of the steel cladding, connections and penetrations. These cut-outs are filled with inserts made of elastic or plastic material. (DG) [de

  4. 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

  5. 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

  6. Transient thermal creep of nuclear reactor pressure vessel type concretes

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1983-01-01

    The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

  7. 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

  8. A three-dimensional rupture analysis of steel liners anchored to concrete pressure and containment vessels

    International Nuclear Information System (INIS)

    Bangash, Y.

    1987-01-01

    Steel liners or plates are anchored to concrete pressure and containment vessels for nuclear and offshore facilities. Due to extreme loading conditions a liner may buckle due to the pull-out or shearing of anchors from the base metal and concrete. Under certain conditions attributed to loadings, liner metal deterioration and cracking of concrete behind the liner, the liner may fail by rupture. This paper presents a three-dimensional analysis of steel-concrete elements, using finite elements analysis in which a provision is made for liner instability, anchor strength and stiffness, concrete cracking and finally liner rupture. The analysis is tested first on an octagonal slab with and without an anchored steel liner. It is then extended to concrete pressure and containment vessels. The analytical results obtained are compared well with those available from the experimental tests and other sources. (author)

  9. 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

  10. 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)

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

    International Nuclear Information System (INIS)

    Naus, D.J.

    1981-05-01

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

  12. A prestressed concrete pressure vessel for helium high temperature reactor system

    International Nuclear Information System (INIS)

    Horner, R.M.W.; Hodzic, A.

    1976-01-01

    A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

  13. Ultimate load analysis of prestressed concrete reactor pressure vessels considering a general material law

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1975-01-01

    A method of analysis is presented, by which progressive fracture processes in axisymmetric prestressed concrete pressure vessels during increasing internal pressure can be evaulated by means of a continuum calculation considering a general material law. Formulations used in the analysis concerning material behaviour are derived on one hand from appropriate results of testing small concrete specimens, and are on the other hand gained by parametric studies in order to solve questions still existing by recalulating fracture tests on concrete bodies with more complex states of stress. Due attention is focussed on investigating the behaviour of construction members subjected to high shear forces (end slabs.). (Auth.)

  14. 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

  15. 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

    The TEMP-STRESS FEM represents an axisymmetric simulation of the reinforced concrete vessel to internal pressurization. The information shows the global deformation, the state of strain/stress within the containment vessel with respect to the imposed pressures. Thus, the location and progress of concrete cracking, the stretching of the liner and the reinforcing bars and final failure are indicated through the entire loading range. Equilibrium of the entire system is assured at definite loading increments. With the progress of concrete cracking, the resisting load is continuously transferred to the reinforcing bars and the liner. Thus, after the tensile strength is exceeded and the concrete stress is set to zero, the internal pressures are entirely resisted by the liner and the reserve strength of the reinforcing bars. The reinforcing bars are mechanically connected to each other by splices, the ultimate strength of which is less than that of the rebars themselves. The corresponding strain at this limiting stress is lower than the ultimate strain of the liner. Therefore, the specified ultimate strength of the splices limits the pressurization of the vessel. Furthermore, once any of the splices fail, then load is transferred to the adjacent members, causing their failure and general failure of the vessel. (orig./HP)

  16. Ultimate Pressure Capacity of Prestressed Concrete Containment Vessels with Steel Fibers

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The ultimate pressure capacity (UPC) of the prestressed concrete containment vessel (PCCV) is very important since the PCCV are final protection to prevent the massive leakage of a radioactive contaminant caused by the severe accident of nuclear power plants (NPPs). The tensile behavior of a concrete is an important factor which influence to the UPC of PCCVs. Hence, nowadays, it is interested that the application of the steel fiber to the PCCVs since that the concrete with steel fiber shows an improved performance in the tensile behavior compared to reinforced concrete (RC). In this study, we performed the UPC analysis of PCCVs with steel fibers corresponding to the different volume ratio of fibers to verify the effectiveness of steel fibers on PCCVs

  17. Experimental study of the structural behavior of the reinforced concrete containment vessel beyond design pressure

    International Nuclear Information System (INIS)

    Oyamada, O.; Saito, H.; Muramatsu, Y.; Hasegawa, T.; Tanaka, N.

    1990-01-01

    The first Advanced Boiling Water Reactor (ABWR) including a reinforced concrete containment vessel (RCCV) is scheduled to be constructed in the 1990s, in Japan. As the RCCV is new to Japan, we performed a trial design, several series of fundamental experiments and partial/total model experiments. This paper presents a summary of the 'TOP SLAB EXPERIMENT' carried out as one of partial model experiments, in which the structural behavior of the RCCV was examined under internal pressure. (orig.)

  18. 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

  19. Tests on model of a prestressed concrete nuclear pressure vessel with multiple cavities

    International Nuclear Information System (INIS)

    Favre, R.; Koprna, M.; Jaccoud, J.P.

    1977-01-01

    The prestressed concrete pressure vessel (prototype) is a cylinder having a diameter of 48 m and a height of 39 m. It has 25 vertical cavities (reactor, heat exchangers, heat recuperators) and 3 horizontal cavities (gas turbines of 500 kw). The cavities are closed by plugs, and their tightness is ensured by a steel lining. A model, on a scale of 1/20, made of microconcrete, was loaded in several cycles, by a uniform inner pressure in the cavities, increasing to the point of failure. The three successive stages were examined: stage of globally elastic behavior, cracking stage, ultimate stage. The behavior of the model is globally elastic up to an inner pressure of 120 to 130 kp/cm 2 , corresponding to about twice the maximum pressure of service, equal to 65 kp/cm 2 . The prestressed tendons at this stage show practically no stress increase. The first detectable cracks appear on the lateral side half-way up the model, as soon as the pressure exceeded 120 kp/cm 2 . From 150-165 kp/cm 2 , the cracking stage can be considered as achieved and the main crack pattern entirely formed. A horizontal crack continues in the middle of the barrel, as well as vertical cracks at each outer cavity. Beyond a pressure of 150-165 kp/cm 2 the ultimate stage begins. The strains of the stresses in the tendons grow more rapidly. The steel lining is highly solicited. Above about 210 kp/cm 2 the model behaves like a structure composed of a group of concrete blocks bound by the tendons and the lining. The failure (240 kp/cm 2 ) occurred through a mechanism of ejection and bending of the concrete ring at the periphery of the barrel of the vessel, which was solicited mainly in tension

  20. 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

  1. 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.

  2. 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

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

    International Nuclear Information System (INIS)

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

    1978-10-01

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

  4. 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

  5. The application of external vibration monitoring to reactors with concrete pressure vessels

    International Nuclear Information System (INIS)

    Hammill, W.J.

    1979-01-01

    The application of external vibration monitoring techniques to advanced gas cooled reactors (AGR) which have concrete pressure vessels is considered. A monitoring system for a particular AGR coolant circuit structure is developed, whose primary objective is to detect impacting of two components, although the detection of forced vibration response is also considered. Experimental results from instrumented components in the reactor and data from rig tests on full size units have been used together with a mathematical model of some elements of the transmission path in order to establish its dynamic characteristics and relate internal component vibration to externally measured signals. The application of external vibration monitoring to the external detection of the forced vibration response of an internal reactor assembly and the remote monitoring of circulator sound output is discussed. (author)

  6. Stress analysis and review of a prestressed concrete reactor pressure vessel of a HTR 500-type

    International Nuclear Information System (INIS)

    Wang, T.J.; Altes, J.

    1988-12-01

    The main aim of this first step of analysis is to test the feasibility of the SMART-code for a complete calculation of PCRV's and to establish the control programs for the finite element analysis. Curved triangular, quadrilateral and membrane shell elements are used. The incremental model in the form of tangential stress-strain law has been chosen as the constitutive model for the concrete. The three parameter failure envelope is used as the failure model of the concrete. For the numerical solution the incremental initial iteration method with constant stiffness is utilized. The creep strain is treated as a liner functional of the stress history and dependent on temperature, humidity and aging. The calculation of the creep behaviour is carried out up to 7 years of operation using the model of SEKI and KAWASUMI. In this model the influence of temperature, humidity and the interaction between them is fully considered. The effects of interaction between temperature and creep with and without humidity's influence are studied and some interesting results are presented. The total creep curves vs time are gained, the deformations of nodal points after 7 years are 1.8 - 5.5 times larger then those of the initial elastic deformation after the first loading. Under the action of prestressing along most parts of the PCRV and under the service condition the main part of the PCRV are in compression. Due to increasing the loading over the operating pressure some parts are cracked and the material behaves nonlinearly. At a loading value of 3.25 times the operating pressure the whole transverse section is fully cracked. For the stage of prestressing, design operating pressure and design limit pressure the vessel behaves elastically. The global safety factory is 1.5 times larger than the design value of 2.25 that shows the conservative design. The analysis method and computer codes, which are used in this review, are confirmed efficiently. (orig./HP)

  7. Overview of experimental results obtained under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Naus, D.J.

    1978-01-01

    Under the Prestressed Concrete Nuclear Pressure Vessel Development Program at the Oak Ridge National Laboratory, various aspects of Prestressed Concrete Pressure Vessels (PCPVs) are investigated and evaluated with respect to reliability, structural performance, constructability, and economy. Based upon identified needs, analytical and experimental investigations are conducted. Areas of interest include finite-element analysis development, materials and structural behavior tests, instrumentation evaluation and development, and structural model tests. Studies have been recently completed in the following areas: concrete embedment instrumentation systems for PCPVs, grouted-nongrouted prestressing systems, acoustic emission as a technique for structural integrity monitoring, and model tests of steam-generator cavity closure plugs for a Gas-Cooled Fast Reactor (GCFR). An overview of results is presented

  8. 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.)

  9. 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)

  10. 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)

  11. Analytical investigation of multicavity prestressed concrete pressure vessels for elastic loading conditions

    International Nuclear Information System (INIS)

    Fanning, D.N.

    1978-09-01

    A three-dimensional finite-element analysis of a commercial high-temperature gas-cooled reactor (HTGR) was made using the finite-element code STATIC-SAP. Four loading conditions were analyzed elastically to evaluate the behavior of the concentric core prestressed concrete reactor vessel (PCRV) of the HTGR. The results of the analysis were evaluated in accordance with Section III, Division 2, of the ASME Code for Reactor Vessels and Containments. The calculated maximum stresses were found to be well within the Code-allowable values. The analysis was preceded by an evaluation of candidate computer codes using comparisons of experimental data with analytical results for the Ohbayashi-Gumi multicavity PCRV model. This vessel was chosen as a basis for comparison because of its geometrical similarity to the large multicavity PCRV and the anticipated availability of a complete set of the original experimental data. The three-dimensional finite-element codes NONSAP and STATIC-SAP were used for the analysis of the Ohbayashi-Gumi vessel

  12. 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.)

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. Structural model testing for prestressed concrete pressure vessels: a study of grouted vs nongrouted posttensioned prestressing tendon systems

    International Nuclear Information System (INIS)

    Naus, D.J.

    1979-04-01

    Nongrouted tendons are predominantly used in this country as the prestressing system for prestressed concrete pressure vessels (PCPVs) because they are more easily surveyed to detect reductions in prestressing level and distress such as results from corrosion. Grouted tendon systems, however, offer advantages which may make them cost-effective for PCPV applications. Literature was reviewed to (1) provide insight on the behavior of grouted tendon system, (2) establish performance histories for structures utilizing grouted tendons, (3) examine corrosion protection procedures for prestressing tendons, (4) identify arguments for and against using grouted tendons, and (5) aid in the development of the experimental investigation. The experimental investigation was divided into four phases: (1) grouted-nongrouted tendon behavior, (2) evaluation of selected new material systems, (3) bench-scale corrosion studies, and (4) preliminary evaluation of acoustic emission techniques for monitoring grouted tendons in PCPVs. The groutability of large tendon systems was also investigated

  20. Compilation of three-dimensional coordinates and specific data of the instrumentation of the prestressed concrete pressure vessel/high temperature helium test rig

    International Nuclear Information System (INIS)

    Klausinger, D.

    1977-04-01

    The positions of the thermoelements, strain gauges of various types, and of Gloetzl instruments installed by SGAE in the model vessel of the Common Project Prestressed Concrete Pressure Vessel/High Temperature Helium Test Rig are defined in cylindrical coordinates. The specific data of the instruments are given like configuration of multiple instruments; type, group and number of the instrument; number of cable and of channel; calibration factors; resistances of instruments and cables. (author)

  1. Multiple shell pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.

    1988-01-01

    A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

  2. 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)

  3. Corrosion of steel tendons in concrete pressure vessels: review of recent literature and experimental investigations

    International Nuclear Information System (INIS)

    Griess, J.C.

    1978-01-01

    The fundamentals of localized corrosion are briefly discussed, and the literature concerning corrosion of carbon steel in aqueous environments, in particular the stress-corrosion cracking of carbon steels, is reviewed. The behavior of high strength steels in specific environments, including concrete and organic substances, is also summarized. The available information indicates that the corrosion of steels in correctly formulated concrete is minimal. Even appreciable concentrations of chloride, sulfate, sulfide, and nitrate salts can be tolerated in the concrete or grout without detrimental effects. Adherence to established standards in the preparation and application of grouts in tendon-bearing conduits should guarantee very long tendon lifetimes. Little is reported about the behavior of tendons in proprietary organic greases or waxes, but very good corrosion resistance is expected if the organic material remains intact. Stress-corrosion cracking tests performed with AISI 1080 steel tendon wires, using the constant-strain-rate method, produced results expected from data in the literature. Cracking was observed only in neutral or acid solutions containing hydrogen sulfide, in ammonium nitrate solutions, and possibly in a dilute solution of sodium bisulfite. General corrosion tests in water and in dilute solutions of sodium nitrate, chloride, or sulfate showed that oxygen was an important factor; corrosion was substantially greater when oxygen had free access to the solution than when access to oxygen was restricted. In the tests with oxygen the heaviest attack on the steel tendons was at the waterline of the solution

  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. Pressure vessel design

    International Nuclear Information System (INIS)

    Annaratone, D.

    2007-01-01

    This book guides through general and fundamental problems of pressure vessel design. It moreover considers problems which seem to be of lower importance but which turn out to be crucial in the design phase. The basic approach is rigorously scientific with a complete theoretical development of the topics treated, but the analysis is always pushed so far as to offer concrete and precise calculation criteria that can be immediately applied to actual designs. This is accomplished through appropriate algorithms that lead to final equations or to characteristic parameters defined through mathematical equations. The first chapter describes how to achieve verification criteria, the second analyzes a few general problems, such as stresses of the membrane in revolution solids and edge effects. The third chapter deals with cylinders under pressure from the inside, while the fourth focuses on cylinders under pressure from the outside. The fifth chapter covers spheres, and the sixth is about all types of heads. Chapter seven discusses different components of particular shape as well as pipes, with special attention to flanges. The eighth chapter discusses the influence of holes, while the ninth is devoted to the influence of supports. Finally, chapter ten illustrates the fundamental criteria regarding fatigue analysis. Besides the unique approach to the entire work, original contributions can be found in most chapters, thanks to the author's numerous publications on the topic and to studies performed ad hoc for this book. (orig.)

  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. Automatic design of prestressed concrete vessels

    International Nuclear Information System (INIS)

    Sotomura, Kentaro; Murazumi, Yasuyuki

    1984-01-01

    Prestressed concrete appeared after high strnegth steel had been produced, therefore it has the history of only 40 years even in Europe where it was developed. High compressive force is given to concrete beforehand by high strength steel to resist tensile force. It is superior to ordinary steel in strength, economy, rust prevention, fire protection and workability, and it competes with ordinary steel in the fields of bridges, towers, water tanks, water pipes, barges, LPG and LNG tanks, reactor pressure vessels, reactor containment vessels and so on. The design of prestressed concrete containment vessels (PCCV) being constructed in Japan adopts the form of mounting a semi-spherical dome on a cylindrical wall of 43m inside diameter and about 1.5m thickness, and the steel pipe sheaths for inserting tendons are arranged in the wall. The Taisei Construction Co. has developed the PC-ADE system which enables the optimum design of PCCVs. The outline of the automatic design system, the design of tendon arrangement, the preparation of the data on the load for stress analysis, the stress analysis by axisymmetric finite element method and the calculation of cross sections are explained. Design is a creative activity, and in the design of PCCVs also, the intention of designers should be materialized when this program is utilized. (Kako, I.)

  8. An overview of experimental results obtained under the prestressed concrete nuclear pressure vessel development program at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Naus, D.J.

    1979-01-01

    Under the Prestressed Concrete Nuclear Pressure Development Program at the Oak Ridge National Laboratory, various aspects of Prestressed Concrete Pressure Vessels (PCPVs) are investigated with respect to reliability, structural performance, constructability, and economy. These investigations are conducted under the High-Temperature Gas-Cooled Reactor (HTGR) Program and the Gas-Cooled Fast Reactor (GCFR) Program. The objectives are to: (1) provide technical support to ongoing PCPV design activities, (2) contribute to the overall technological data base, and (3) provide independent review and evaluations. Specific areas of interest at present include finite-element analysis development, materials and structural behaviour tests, instrumentation evaluations and development, and structural model tests. The following provides an overview of both the HTGR and GCFR PCPV activities and a summary of recent experimental results

  9. 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.)

  10. Experimental investigation of stresses and deformations of the model of a pod-boiler-prestressed concrete pressure vessel. Pt. 1

    International Nuclear Information System (INIS)

    Stoever, R.

    1973-01-01

    Investigations of elastic models are suitable to obtain independent values for stress states and deformations of thickwalled pressure vessels to check computer programs for three-dimensional elastic calculations. An elastic model of epoxy resin was constructed with the geometry of the pod boiler pressure vessel of the Hartlepool nuclear power station. With this model strains and deformations were measured for internal pressure. The stress states in the neighbourhood of the large vertical openings for the boiler pods and the horizontal gas ducts and at the junction of cylinder and plates were of special interest. Therefore most of the gauges were concentrated in these regions. A considerable number of strain gauges were embedded in the wall. The construction of the model is described in part one and results of the measurements are presented and discussed in part two of this report. (orig.) [de

  11. 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

  12. 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

  13. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1977-01-01

    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. (Auth.)

  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. Reactor pressure vessel design

    International Nuclear Information System (INIS)

    Foehl, J.

    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 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

  16. Stress analysis of liners for prestressed concrete reactor pressure vessels with regard to non-linear behaviour of liner material and of anchor-characteristics

    International Nuclear Information System (INIS)

    Oberpichler, R.; Schnellenbach, G.

    1975-01-01

    The thin liner attached by anchors like a membrane to the interior wall of a prestressed concrete reactor pressure vessel (PCRV) has to provide the leak-tightness of the vessel. Furthermore the liner may serve as internal shuttering for placing of concrete as well as a support for the cooling system. The two-dimensional behaviour of the liner is investigated with regard to non-linear anchor-characteristics and non-linear material behaviour of the liner. The analysis is based on a plane stress model under the assumption of a membrane state of the liner. Calculations are performed by the dynamic relaxation method. With the aid of available non-linear stress-strain diagrams, describing the post-buckling behaviour, individual panels are considered as buckled ones. The adjacent unbuckled panels are calculated on other non-linear diagrams. Strains and stresses in the liner and additional shear loads in the anchors can be calculated with arbitrary sizing and spacing of the anchors. With respect to the parameters they are easily controlled. Since actual loads on the liner are defined by the PCRV-behaviour, an economical and safe design is possible. Finally an extreme case is calculated to assess the maximum value of the shear-forces assuming zero post-buckling capacity for the buckled panel. (Auth.)

  17. 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.

  18. 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

  19. 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)

  20. Concrete containment vessels (CCV) for nuclear power plants, (1)

    International Nuclear Information System (INIS)

    Ibe, Yukimi; Kitajima, Masatake

    1977-01-01

    Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

  1. 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)

  2. 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

  3. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. 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.

  5. 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

  6. Pressure vessel lid

    International Nuclear Information System (INIS)

    Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

    1986-01-01

    The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

  7. Review of concrete properties for prestressed concrete pressure vesssels

    International Nuclear Information System (INIS)

    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 0 C (212 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. Prestressed concrete vessels suitable for helium high temperature reactors

    International Nuclear Information System (INIS)

    Lockett, G.E.; Kinkead, A.N.

    1967-02-01

    In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

  9. Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

    1986-01-01

    A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

  10. 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)

  11. 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

  12. 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

  13. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1977-01-01

    Several present and proposed gas-cooled reactors use concrete pressure vessels. In addition, concrete is almost universally used for the secondary containment structures of water-cooled reactors. Regulatory agencies must have means of assuring that these concrete structures perform their containment functions during normal operation and after extreme conditions of transient overpressure and high temperature. The NONSAP nonlinear structural analysis program has been extensively modified to provide one analytical means of assessing the safety of reinforced concrete pressure vessels and containments. Several structural analysis codes were studied to evaluate their ability to model the nonlinear static and dynamic behavior of three-dimensional structures. The NONSAP code was selected because of its availability and because of the ease with which it can be modified. In particular, the modular structure of this code allows ready addition of specialized material models. Major modifications have been the development of pre- and post-processors for mesh generation and graphics, the addition of an out-of-core solver, and the addition of constitutive models for reinforced concrete subject to either long-term or short-term loads. Emphasis was placed on development of a three-dimensional analysis capability

  14. 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)

  15. 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.)

  16. The pressure vessel for the NSF tandem

    International Nuclear Information System (INIS)

    Jones, C.W.

    1979-04-01

    The pressure vessel is a major component of the 30 MV tandem Van de Graaff electrostatic accelerator to be used in nuclear structure research at Daresbury Laboratory. The accelerator will be capable of accelerating the full range of ions in the form of a beam. Acceleration takes place in a vertical evacuated tube (beam tube) by means of a high potential on a terminal at the central position, the terminal and beam tube assembly being supported by an insulated stack structure within the pressure vessel. Under operating conditions the vessel is filled with sulphur hexafluoride gas (SF 6 ) at high pressure which acts as an insulating medium between the centre terminal and the vessel wall. The vessel is situated inside a concrete tower which besides supporting the injector room above the vessel also acts as radiation shielding around the accelerator. The report covers: functional requirements; fundamental considerations with regard to the design and procurement; detail design; materials; manufacture; acceptance test; surface treatment; final leak test. (U.K.)

  17. 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.

  18. Mark III Containment vessel/annulus concrete design

    International Nuclear Information System (INIS)

    Chang, P.S.; Moussa, M.M.

    1981-01-01

    Recently, engineers have been considering the significant dynamic impact of safety/relief valve (S/RV) discharge loads on the containment structures, safety equipment, and piping systems in BWR type reactors. For a plant in the construction stage, extensive modifications will be made to qualify these new loads. The lower portion of the containment vessel serves as a suppression pool pressure boundary and is designed to sustain the effects of postulated loss of coolant accidents, seismic occurrences, S/RV discharge loads, and other effects. Extremely high spectral peak accelerations of the free-standing steel containment vessel can be obtained during the air dearing process of the S/RV discharge. Parametric studies indicated that a substantial reduction in response can be obtained by increasing the stiffness of the steel containment vessel in the lover area. A concrete backing configuration in the suppression pool area of Mark III Containment is proposed in this paper. A composite action is assumed between the steel containment vessel shell and the concrete section. The system is physically separated from the shield building. This approach warrants an early erection of the shield building and a late installation of piping systems in the containment vessel suppression pool area. Finite element analyses are performed by using ASHSD2 and EASE2 computer codes. The results of the analyses have shown the proposed stress criteria are satisfied. The approach pressented is justified to be a workable system for a new plant design. (orig./HP)

  19. 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

  20. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  1. 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

  2. Closure system of a vessel made of prestressed concrete

    International Nuclear Information System (INIS)

    Audibert, Alain

    1974-01-01

    The present invention relates to removable plugs of prestressed concrete which can be fitted to every type of closed high pressure vessels and especially to the cylindrical vessels of nuclear reactors. The method involved permits the plug to be fitted to the vessel through both radial and axial prestress. In this purpose, said invention proposes removable prestress ribs fitted inside sheaths in the plug and extending throughout the upper part of the bearing surfaces of the plug, said ribs being regularly arranged along the generators of an hyperboloid of one sheet. Owing to this important feature, that is to say said inclination of the ribs in accordance with the generators of said hyperboloid, said rib inclination can be changed on requirement for each realization [fr

  3. 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

  4. 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

  5. Design criteria for high-temperature-affected, metallic and ceramic components, and for the prestressed concrete reactor pressure vessel of future HTR systems. Final report. Vol. 1-4

    International Nuclear Information System (INIS)

    1988-08-01

    This work in five separate volumes reports on the elaboration of basic data for the formulation of design criteria for HTR components and is arranged into the four following subject areas : (1) safety-specific limiting conditions; (2) metallic components; (3) prestressed concrete reactor pressure vessels; (4) graphitic reactor internals. Under item 2, the mechanical and physical characteristics of the materials X20CrMoV 12 1, X10NiCrAlTi 32 20, and NiCr23Co12Mo are examined up to temperatures of 950deg C. Stress-strain rate laws are elaborated for description of the inelastic deformation behavior. The representation of the subject area reactor pressure vessels deals with four main topics: Prestressed concrete support structure, liner, vessel closures, thermal protection system. Quality-assurance classes are defined under item 4 for graphitic components and load levels for load categories. The material evaluation is discussed in detail (e.g. manufacturing monitoring from the raw material to the graphitization and manufacturing testing up to the acceptance test). In addition, the corrosion behavior and irradiation behavior of graphite is examined and rules for computation of stresses in irradiated and unirradiated graphitic components are elaborated. (MM) [de

  6. 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)

  7. 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

  8. 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.

  9. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1985-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

  10. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1980-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

  11. Behaviour of a pre-stressed concrete pressure-vessel subjected to a high temperature gradient; Comportement d'un caisson en beton precontraint soumis a un gradient de temperature eleve

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Bonvalet, Ch; Dawance, G; Marechal, J C [Centre Experimental de Recherches et d' Etudes du Batiment et des Travaux Publics (CEBTP), 76 - Harfleur (France)

    1965-07-01

    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{sup 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

  12. Behaviour of a pre-stressed concrete pressure-vessel subjected to a high temperature gradient; Comportement d'un caisson en beton precontraint soumis a un gradient de temperature eleve

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, F. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Bonvalet, Ch.; Dawance, G.; Marechal, J.C. [Centre Experimental de Recherches et d' Etudes du Batiment et des Travaux Publics (CEBTP), 76 - Harfleur (France)

    1965-07-01

    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{sup 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

  13. Radiation resistance of concrete of nuclear reactor vessel

    International Nuclear Information System (INIS)

    Belyakov, V.V.; Denisov, A.V.; Korenevskij, V.V.; Muzalevskij, L.P.; Dubrovskij, V.B.; Ivanov, D.A.; Nazarov, I.L.; Sashin, N.L.

    1992-01-01

    Results of calculational-experimental determination of radiation resistance for concrete bases on limestone gravel and quartz sand, which are the most perspective materials for manufacturing prestressed concrete of the VG-400 reactor vessel are considered. Material samples under investigation were irradiated in the channels of the IBR-2 research reactor for the purpose of the calcultional result verification

  14. Review of analysis methods for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Dodge, W.G.; Bazant, Z.P.; Gallagher, R.H.

    1977-02-01

    Theoretical and practical aspects of analytical models and numerical procedures for detailed analysis of prestressed concrete reactor vessels are reviewed. Constitutive models and numerical algorithms for time-dependent and nonlinear response of concrete and various methods for modeling crack propagation are discussed. Published comparisons between experimental and theoretical results are used to assess the accuracy of these analytical methods

  15. Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

    International Nuclear Information System (INIS)

    Bodmann, E.

    1984-01-01

    Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

  16. Nuclear power plant pressure vessels. Inservice inspections

    International Nuclear Information System (INIS)

    1995-01-01

    The requirements for the planning and reporting of inservice inspections of nuclear power plant pressure vessels are presented. The guide specifically applies to inservice inspections of Safety class 1 and 2 nuclear power plant pressure vessels, piping, pumps and valves plus their supports and reactor pressure vessel internals by non- destructive examination methods (NDE). Inservice inspections according to the Pressure Vessel Degree (549/73) are discussed separately in the guide YVL 3.0. (4 refs.)

  17. Behavior of cracked concrete nuclear containment vessels during earthquakes

    International Nuclear Information System (INIS)

    Gergely, P.; Stanton, J.F.; White, R.N.

    1975-01-01

    When pressure builds up in a reinforced concrete nuclear containment shell, its cylindrical wall cracks vertically and horizontally at intervals of about five feet. If an earthquake occurs simultaneously with this pressurization, inertia forces are transmitted across the horizontal crack planes. The forces and deformations must be small enough to maintain the integrity of the steel liner. A typical containment shell has a radius of about 65 ft. and a wall thickness of about 4 ft. It is heavily reinforced with vertical, horizontal, and sometimes diagonal bars. A steel shell of about 3 / 8 in. thickness is attached to the concrete with anchors. The seismic shear forces are transmitted across the horizontal cracks by interface shear transfer (combination of shear friction and aggregate interlocking), by dowel action of the bars, and by diagonal bars if they are used. One important question in the design of such vessels is whether the diagonal bars are necessary. In the experimental portion of the current investigation several types of tests were conducted to study the load-slip characteristics of interface shear transfer under high intensity cyclic loading. In some cases external bars provided the clamping action of reinforcement, in more recent tests large diameter embedded bars were used. This presentation summarizes the analytical part of the investigation. A representative load-slip curve has been used in the analyses to assess the intensity of the stresses and deformations, and to study the importance of the variables as an aid in planning future tests

  18. 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

  19. Milestones in pressure vessel technology

    International Nuclear Information System (INIS)

    Spence, J.; Nash, D.H.

    2004-01-01

    The progress of pressure vessel technology over the years has been influenced by many important events. This paper identifies a number of 'milestones' which have provided a stimulus to analysis methods, manufacturing, operational processes and new pressure equipment. The formation of a milestone itself along with its subsequent development is often critically dependent on the work of many individuals. It is postulated that such developments takes place in cycles, namely, an initial idea, followed sometimes by unexpected failures, which in turn stimulate analysis or investigation, and when confidence is established, followed finally by the emergence of codes ad standards. Starting from the industrial revolution, key milestones are traced through to the present day and beyond

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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

  5. 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

  6. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1979-01-01

    The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

  7. 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.

  8. The instrumentation of the prestressed concrete vessel with hot liner at Seibersdorf Research Centre

    International Nuclear Information System (INIS)

    Zemann, H.

    1975-11-01

    The joint project ''Prestressed Concrete Pressure Vessel with Hot Liner'' at Seibersdorf Research Centre now is in the process of testing the PCPV both in construction and operation from the safety point of view. The physical state of the PCPV (modulus of elasticity, humidity of concrete, creeping, etc.) is brought to stable conditions by ''pre-aging''. In order to control this process of stabilisation, an extensive knowledge of the concrete and an elaborated instrumentation is a necessity. This paper presents a survey about the philosophy and the realisation of the instrumentation of the PCPV and the investigations we performed to interpret the measurements. (author)

  9. Prestressed concrete reactor vessels: review of design and failure criteria

    International Nuclear Information System (INIS)

    Endebrock, E.G.

    1975-03-01

    The design and failure criteria of prestressed concrete reactor vessels (PCRVs) are reviewed along with the analysis methods. The mechanical properties of concrete under multiaxial stresses are not adequately quantified or described to permit an accurate analysis of a PCRV. Structural analysis of PCRVs almost universally utilizes a finite element which encounters difficulties in numerical solution of the governing equations and in treatment of fractured elements. (U.S.)

  10. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  11. Hydrogen storage in insulated pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  12. Development of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

  13. Development of PWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

  14. Expanded Fermilab pressure vessel directory program

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

  15. Expanded Fermilab pressure vessel directory program

    International Nuclear Information System (INIS)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

  16. 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)

  17. 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

  18. Ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Haniger, L.

    1980-01-01

    Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

  19. Reactor pressure vessel. Status report

    International Nuclear Information System (INIS)

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

    1996-10-01

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

  20. Metallurgy of steels for PWR pressure vessels

    International Nuclear Information System (INIS)

    Kepka, M.; Mocek, J.; Barackova, L.

    1980-01-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

  1. Metallurgy of steels for PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

    1980-09-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

  2. Some aspects of reactor pressure vessel integrity

    International Nuclear Information System (INIS)

    Korosec, D.; Vojvodic, G.J.

    1996-01-01

    Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

  3. 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.)

  4. Nuclear reactor pressure vessel flaw distribution development

    International Nuclear Information System (INIS)

    Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

    1991-12-01

    Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

  5. 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.)

  6. 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)

  7. Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.; D'Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

    1993-09-01

    The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The 'Concrete Benchmark' experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the 'Concrete Benchmark' experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs

  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. 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

  10. Behaviour of concrete containment under over-pressure conditions

    International Nuclear Information System (INIS)

    Atchison, R.J.; Asmis, G.J.K.; Campbell, F.R.

    1979-01-01

    The Atomic Energy Control Board of Canada initiated June, 1975, a major study of the behaviour of concrete containment under over-pressure conditions. Although extensive theoretical and experimental work has been carried out for thick-walled Prestressed Concrete Reactor Vessels (PCRV's), there is a want of information on the non-linear response of thin-walled structures typical of the CANDU, 600 MW(e) cylindrical/spherical, post-tensioned containment shells. The purpose of this paper is to provide an overview of the total program, to present the reasons behind the research contract, and the specification and implementation of the work. The results of the theoretical and experimental work and their implications with respect to Canadian Concrete Containment practice are discussed. This study is unique, and, as far as is known, has no world-wide precedence. (orig.)

  11. 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....

  12. Increase of cyclic durability of pressure vessels

    International Nuclear Information System (INIS)

    Vorona, V.A.; Zvezdin, Yu.I.

    1980-01-01

    The durability of multilayer pressure vessels under cyclic loading is compared with single-layer vessels. The relative conditional durability is calculated taking into account the assumption on the consequent destruction of layers and viewing a vessel wall as an indefinite plate. It is established that the durability is mainly determined by the number of layers and to a lesser degree depends on the relative size of the defect for the given layer thickness. The advantage of the multilayer vessels is the possibility of selecting layer materials so that to exclude the effect of agressive corrosion media on the strength [ru

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-01-01

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

  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. Prediction of moisture migration and pore pressure build-up in concrete at high temperatures

    International Nuclear Information System (INIS)

    Ichikawa, Y.; England, G.L.

    2004-01-01

    Prediction of moisture migration and pore pressure build-up in non-uniformly heated concrete is important for safe operation of concrete containment vessels in nuclear power reactors and for assessing the behaviour of fire-exposed concrete structures. (1) Changes in moisture content distribution in a concrete containment vessel during long-term operation should be investigated, since the durability and radiation shielding ability of concrete are strongly influenced by its moisture content. (2) The pressure build-up in a concrete containment vessel in a postulated accident should be evaluated in order to determine whether a venting system is necessary between liner and concrete to relieve the pore pressure. (3) When concrete is subjected to rapid heating during a fire, the concrete can suffer from spalling due to pressure build-up in the concrete pores. This paper presents a mathematical and computational model for predicting changes in temperature, moisture content and pore pressure in concrete at elevated temperatures. A pair of differential equations for one-dimensional heat and moisture transfer in concrete are derived from the conservation of energy and mass, and take into account the temperature-dependent release of gel water and chemically bound water due to dehydration. These equations are numerically solved by the finite difference method. In the numerical analysis, the pressure, density and dynamic viscosity of water in the concrete pores are calculated explicitly from a set of formulated equations. The numerical analysis results are compared with two different sets of experimental data: (a) long-term (531 days) moisture migration test under a steady-state temperature of 200 deg. C, and (b) short-term (114 min) pressure build-up test under transient heating. These experiments were performed to investigate the moisture migration and pressure build-up in the concrete wall of a reactor containment vessel at high temperatures. The former experiment simulated

  16. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Science.gov (United States)

    2010-10-01

    ... that each pressure vessel, including each volume tank, cylinder and PVHO, and each pressure piping... tests conducted in accordance with this section shall be either hydrostatic tests or pneumatic tests. (1... times the maximum allowable working pressure. (2) When a pneumatic test is conducted on a pressure...

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

    Energy Technology Data Exchange (ETDEWEB)

    Lafitte, R.; Marchand, J. D. [Bonnard et Gardel, Ingenieurs-Conseil, Lausanne (Switzerland)

    1981-01-15

    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.

  18. 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

  19. Flexible Composite-Material Pressure Vessel

    Science.gov (United States)

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

    2003-01-01

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

  20. Procurement of replacement pressure vessels for MURR

    International Nuclear Information System (INIS)

    Meyer, W.A. Jr.; Edwards, C.B. Jr.; McKibben, J.C.; Schoone, A.R.

    1989-01-01

    The University of Missouri Research Reactor Facility (MURR) located in Columbia, Missouri, is the highest powered, highest steady-state flux university research reactor in the United States. The reactor is a 10-MW pressurized loop, in-pool-type, light-water-moderated, beryllium-reflected, flux trap reactor. MURR has a compact core (0.033 m 3 ) composed of eight fuel elements of the materials test reactor type arranged as an annular right circular cylinder between the inner and outer aluminum pressure vessels. Conservative engineering judgment resulted in the decision in 1988 to purchase new inner and outer pressure vessels. This paper details the difficulties encountered in procuring replacements for aluminum pressure vessels built to standards that are no longer applicable in attempting to meet nuclear standards that are not applicable to nonferrous material

  1. Nickel hydrogen common pressure vessel battery development

    Science.gov (United States)

    Jones, Kenneth R.; Zagrodnik, Jeffrey P.

    1992-01-01

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

  2. Research and development of the prestressed concrete reactor vessel

    International Nuclear Information System (INIS)

    Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

    1975-01-01

    Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

  3. Head spray nozzle in reactor pressure vessel

    International Nuclear Information System (INIS)

    Hatano, Shun-ichi.

    1990-01-01

    In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

  4. Foundamental characteristics of layered pressure vessel

    International Nuclear Information System (INIS)

    Moriwaki, Yoshikazu; Fugino, Masayuki; Shimizu, Yasuhiro; Nakamura, Takeshi

    1978-01-01

    Pressure vessels become larger and the working pressure become higher with the remarkable development of petroleum, chemical, thermal power generation and atomic energy industries. Multi-layered pressure vessels can be manufactured cheaply without large installations, and large wall thickness can be made, therefore they are suitable for large pressure vessels. The stress and deformation behaviors of such vessels are very complex because of the effect of frictional force working between layers. In this study, the phenomena arising in multiple layers and the difference as compared with single wall were studied fundamentally as one step for analyzing multi-layered pressure vessels as a whole. Finite element technique was employed as the analyzing method, and the behavior of multiple layers was analyzed, regarding it as multiple contact problem. The behavior of multiple layers seems to appear conspicuously in case of bending load, therefore the basic characteristics regarding bending were examined. The evaluation of interfacial stiffness was carried out by experiment. The computer program for analyzing multiple contact problem was developed. In order to examine the validity of the program, comparison with the analytical solution heretofore and the result of calculation by finite element technique was carried out. Moreover, the experimental proof with multi-layered models was made. The frictional force between layers hardly contributes to the stiffness. (Kako, I.)

  5. 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

  6. 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)

  7. 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

  8. Crack propagation on spherical pressure vessels

    International Nuclear Information System (INIS)

    Lebey, J.; Roche, R.

    1975-01-01

    The risk presented by a crack on a pressure vessel built with a ductile steel cannot be well evaluated by simple application of the rules of Linear Elastic Fracture Mechanics, which only apply to brittle materials. Tests were carried out on spherical vessels of three different scales built with the same steel. Cracks of different length were machined through the vessel wall. From the results obtained, crack initiation stress (beginning of stable propagation) and instable propagation stress may be plotted against the lengths of these cracks. For small and medium size, subject to ductile fracture, the resulting curves are identical, and may be used for ductile fracture prediction. Brittle rupture was observed on larger vessels and crack propagation occurred at lower stress level. Preceedings curves are not usable for fracture analysis. Ultimate pressure can be computed with a good accuracy by using equivalent energy toughness, Ksub(1cd), characteristic of the metal plates. Satisfactory measurements have been obtained on thin samples. The risks of brittle fracture may then judged by comparing Ksub(1cd) with the calculated K 1 value, in which corrections for vessel shape are taken into account. It is thus possible to establish the bursting pressure of cracked spherical vessels, with the help of two rules, one for brittle fracture, the other for ductile instability. A practical method is proposed on the basis of the work reported here

  9. Pressure vessel failure at high internal pressure

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1995-01-01

    A RPV failure due to plastic instability was investigated using the ABAQUS finite element code together with a material model of thermal plasticity for large deformations. Not only rotational symmetric temperature distributions were studied, but also 'hot spots'. Calculations show that merely by the depletion of strength of the material - even at internal wall temperatures well below the melting point of the fuel elements of about 2000/2400 C - the critical internal pressure can decrease to values smaller than the operational pressure of 16 Mpa. (orig.)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  11. 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

  12. 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)

  13. 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

  14. 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

  15. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1981-01-01

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  16. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  17. Problems in Pressure Vessel Design and Manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Hellstroem, O [Uddeholms AB, Degerfors (Sweden); Nilson, Ragnar [AB Atomenergi, Nykoeping (Sweden)

    1963-05-15

    The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels.

  18. Problems in Pressure Vessel Design and Manufacture

    International Nuclear Information System (INIS)

    Hellstroem, O.; Nilson, Ragnar

    1963-05-01

    The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels

  19. 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.)

  20. Guidelines for pressure vessel safety assessment

    Science.gov (United States)

    Yukawa, S.

    1990-04-01

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

  1. Pressure vessel for gaseous media

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-07-14

    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/sup 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.

  2. 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)

  3. 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

  4. 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)

  5. Energy and impacts of pressure vessel explosions

    International Nuclear Information System (INIS)

    Kurttila, H.

    1999-01-01

    In this paper the explosion energy is considered to be same as the energy of pressure vessel discharge. This is the maximum energy which can be obtained from the process. The energy can be used or it can cause the violence of an explosion accident. (orig.)

  6. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1977-01-01

    This paper discusses the development of a finite element code suitable for the safety analysis of prestressed concrete reactor vessels. The project has involved modification of a general purpose computer code to handle reinforced concrete structures as well as comparison of results obtained with the code against published experimental data. The NONSAP nonlinear structural analysis program was selected for the ease with which it can be modified to encompass problems peculiar to nuclear reactors. Pre- and post-processors have been developed for mesh generation and for graphical display of response variables. An out-of-core assembler and solver have been developed for the analysis of large three dimensional problems. The constitutive model for short term loads forms an orthotropic stress-strain relationship in which the concrete and the reinforcing steel are treated as a composite. The variation of stiffness and strength of concrete under multiaxial stress states is accounted for. Cracks are allowed to form at element integration points based on a three dimensional failure envelope in stress space. Composite tensile and shear properties across a crack are modified to account for bond degradation and for dowel action of the reinforcement. The constitutive law for creep is base on the expansion of the usual creep compliance function in the form of a Dirichlet exponential series. Empirical creep data are then fit to the Dirichlet series approximation by means of a least squares procedure. The incremental deformation process is subsequently reduced to a series of variable stiffness elasticity problems in which the past stress history is represented by a finite number of hidden material variables

  7. 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

  8. Eddy current testing of composite pressure vessels

    Science.gov (United States)

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

    2018-04-01

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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. A framework expert system for pressure vessels

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. 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.)

  15. 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.

  16. 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.

  17. 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)

  18. 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.

    1978-01-01

    The variable modulus-cracking model is capable of predicting the behavior of reinforced concrete structures (such as the reinforced plate under transverse pressure described previously) well into the range of nonlinear behavior including the prediction of the ultimate load. For unreinforced thick-walled concrete vessels under internal pressure the use of elastic--plastic concrete models in finite element codes enhances the apparent ductility of the vessels in contrast to variable modulus-cracking models that predict nearly instantaneous rupture whenever the tensile strength at the inner wall is exceeded. For unreinforced thick-walled end slabs representative of PCRV heads, the behavior predicted by finite element codes using variable modulus-cracking models is much stiffer in the nonlinear range than that observed experimentally. Although the shear type failures and crack patterns that are observed experimentally are predicted by such concrete models, the ultimate load carrying capacity and vessel-ductility are significantly underestimated. It appears that such models do not adequately model such features as aggregate interlock that could lead to an enhanced vessel reserve strength and ductility

  19. 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)

  20. The design of bonded reinforcement for thermal stresses in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Kotulla, B.; Hansson, V.

    1977-01-01

    This paper deals with examples of thermal loadings where instationary growth of tensile zones and redistribution of stresses by cracking are of importance. Temperatures produce, in addition to prestressing and internal pressure, the most important stresses in a prestressed concrete reactor pressure vessel. Characteristic of thermal stresses is that they are influenced to a large extent by creep of concrete and that they influence stress redistributions by temperature dependent creep data. Computations show that during the first instationary heating process of the vessel stresses are reduced by creep effects to about fifty percent of the values of the stationary elastic case at the hot face. With a following cooling, creep effects are generally much less, so this case may produce tensile stresses on the internal face of the wall which lead to cracking of the concrete. Tensile stresses first occur due to the instationary growth of the temperature field in a narrow zone near the liner. If outside this zone compressive stresses exist due to prestressing then crack spreading is limited and restraint by the parts of the wall under compression causes crack distribution even without reinforcement in this zone. Growth of cracks with the instationary spreading of tensile zones according to temperature development was calculated. These calculations take into account discrete cracks, reinforcement and different assumptions for tensile strength. Reinforcement of small diameter near the surface has the best influence on crack spacing. Calculations show that for the stationary state of cooling the forces in the reinforcement may be as low as twenty to thirty percent of the tensile force not taking into account cracking of the concrete

  1. 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.)

  2. 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.

  3. 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

  4. 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)

  5. 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)

  6. The Use of Prestressed Concrete Vessels in the French Power Reactor Programme

    International Nuclear Information System (INIS)

    Conte, F.; Dambrine, C.; Gaussot, D.

    1963-01-01

    This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author) [fr

  7. Modeling Scala Media as a Pressure Vessel

    Science.gov (United States)

    Lepage, Eric; Olofsson, A.˚Ke

    2011-11-01

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

  8. 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...

  9. 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.)

  10. 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

  11. 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)

  12. 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.)

  13. Prestressed cast iron pressure vessels as burst-proof pressure vessels for innovative nuclear applications

    International Nuclear Information System (INIS)

    Froehling, W.; Boettcher, A.; Bounin, D.; Steinwarz, W.; Geiss, M.; Trauth, M.

    2000-01-01

    The amendment to the German Atomic Energy Act from July 28, 1994 requires that events 'whose occurrence is practically excluded by the measures against damages', i.e. events of the category residual risk, must not necessitate far reaching protective measures outside the plant. For a conventional reactor pressure vessel, the residual risk consists in the very small probability of a catastrophic failure (formation of a large fracture opening, bursting of the vessel). With a prestressed cast iron vessel (PCIV), the formation of a large fracture opening or bursting of the vessel, respectively, is impossible due to its design properties. Against this background the possibility of the use of this type of pressure vessel for lightwater reactors has been studied in the frame of a 'Working Group for Innovative Nuclear Technology', founded by different research institutes and industrial companies. Furthermore, it has been studied whether the use of the PCIV support the realization of a corecatcher system. The results are presented in this report. Already many years earlier, Siempelkamp has performed industrial development and Forschungszentrum Juelich related experimental and theoretical safety research for the PCIV as an innovative, bust-proof pressure vessel concept. This development of the PCIV as well as its safety properties are also presented in a conclusive manner. (orig.) [de

  14. Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

    International Nuclear Information System (INIS)

    Larson, G.C.

    1980-01-01

    This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

  15. 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)

  16. Design of pressure vessels. Part 1

    International Nuclear Information System (INIS)

    Grandemange, J.M.

    2008-01-01

    The equipments and loops of PWR reactors are basically pressure vessels. Their specificities concern the integrity warranties that must be implemented considering their importance for the reactors safety. Thus, stress is put on the exhaustiveness of the prevention of in-service degradation and on the safety scenarios considered. The second specificity concerns the possibility of activation of wear and corrosion products during their flow inside the reactor core. This second aspect leads to some constraints on the choice of the materials used and on the surface coating of the inside wall of big components of the primary circuit. The aim of this document is to develop the general approach adopted for the design of the pressure vessels of PWR fluid loops, and to stress more particularly on the nuclear particularities of these equipments. Some extensions of these rules to high temperature resistant materials (FBR-type reactors) are also evoked. Content: General considerations: design basis of pressure vessels, risk analysis and design conditions, ruining paths and safety coefficients; 2 - damage prevention for excessive deformation: definitions, criteria; 3 - prevention of the plastic instability damage: definition, criteria; 4 - buckling prevention: definition and mechanisms, rules and criteria; 5 - prevention of progressive deformation damage: definitions, plastic adaptation, plastic accommodation, progressive deformation; 6 - prevention of fatigue damage: definitions, general prevention approach, design fatigue curves, analytic approach, particular aspects, analysis of zones with geometrical singularity; 7 - prevention of sudden rupture damage: fragile rupture and ductile tear, general approach, analytic criteria, irradiation and aging effects; 8 - other potential damages; 9 - conclusion. (J.S.)

  17. 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

  18. Pressure Vessel Steel Research: Belgian Activities

    International Nuclear Information System (INIS)

    Van Walle, E.; Fabry, A.; Ait Abderrahim, H.; Chaouadi, R.; D'hondt, P.; Puzzolante, J.L.; Van de Velde, J.; Van Ransbeeck, T.; Gerard, R.

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly

  19. Pressure Vessel Steel Research: Belgian Activities

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E; Fabry, A; Ait Abderrahim, H; Chaouadi, R; D` hondt, P; Puzzolante, J L; Van de Velde, J; Van Ransbeeck, T [Centre d` Etude de l` Energie Nucleaire, Mol (Belgium); Gerard, R [TRACTEBEL, Brussels (Belgium)

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly.

  20. 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)

  1. 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

  2. Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

    International Nuclear Information System (INIS)

    Loeb, Andreas; Stanke, Dieter; Thoma, Markus; Eisenmann, Beata; Prechtl, Erwin; Dehnke, Burckhard

    2008-01-01

    The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

  3. 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)

  4. 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

  5. Midland reactor pressure vessel flaw distribution

    International Nuclear Information System (INIS)

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

    1993-12-01

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

  6. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    Foerli, O.; Hernes, T.

    1978-01-01

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

  7. 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.)

  8. 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)

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  14. Safety margin evaluation of pre-stressed concrete nuclear containment vessel model with BARC code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Patnaik, R.; Ramanujam, S.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian pressurised heavy water reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results and for prediction of safety margins of Indian PHWRs. The present paper highlights the analysis results for prestressed concrete containment vessel (PCCV) tested at Sandia National Labs, USA in a round robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd = design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd

  15. 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

  16. Pressure thermal shock analysis for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Galik, G.; Kutis, V.; Jakubec, J.; Paulech, J.; Murin, J.

    2015-01-01

    The appearance of structural weaknesses within the reactor pressure vessel or its structural failure caused by crack formation during pressure thermal shock processes pose as a severe environmental hazard. Coolant mixing during ECC cold water injection was simulated in a detailed CFD analysis. The temperature distribution acting on the pipe wall internal surface was calculated. Although, the results show the formation of high temperature differences and intense gradients, an additional structural analysis is required to determine the possibility of structural damage from PTS. Such an analysis will be the subject of follow-up research. (authors)

  17. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H; Ahlstroem, P E; Pershagen, B

    1961-04-15

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D{sub 2}O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960.

  18. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

    1961-04-01

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D 2 O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  19. 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)

  20. Temperature field in the bottom of concrete reactor vessel; Temperaturno polje u podu betonskog reaktorskog suda

    Energy Technology Data Exchange (ETDEWEB)

    Jovasevic, V; Tosic, D; Zaric, S; Maksimovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1969-07-01

    This paper contains detailed scheme of reactor bottom vessel made of concrete and the results of calculated relevant temperature distribution. Method applied for calculation is described taking into account all relevant factors and assuming that thermal conductivity of concrete is homogeneous and independent of temperature.

  1. Apparatus for carrying out ultrasonic inspection of pressure vessels

    International Nuclear Information System (INIS)

    Dent, K.H.; Greenhalgh, F.G.

    1975-01-01

    An apparatus is described for moving an ultrasonic scanning mechanism over the interior surface of a pressure vessel and comprising a mast for supporting the scanning mechanism inside the vessel and a carriage for traversing the mast within the vessel, the mast being pivotably secured to the carriage so that when the ultrasonic scanning mechanism contacts the interior surface of the pressure vessel the mast is caused to pivot. (auth)

  2. 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

  3. Dictionary of pressure vessel and piping technology

    International Nuclear Information System (INIS)

    Schmitz, H.P.

    1987-01-01

    This dictionary is the result of many years of evaluation of technical terminology taken from the salient non-German rules, regulations, standards and specifications such as ANSI, API, ASME, ASNT, ASTM, BSI, EJMA, TEMA, and WRC (see bibliography) and of comparing these with the corresponding German rules, regulations, etc., as well as examining relevant technical documentation. This dictionary fills the gap left by existing dictionaries. The following specialized factors are given special attention: pressure vessels, tanks, heat exchangers, piping, valves and fittings, expansion joints, flanges, giving particular consideration to the fields of materials, welding, strength calculation, design and construction, fracture mechanics, destructive and non-destructive testing, as well as heat and mass transfer. (orig./HP) [de

  4. 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

  5. Acoustic emission monitoring of a pressure vessel

    International Nuclear Information System (INIS)

    Birchon, D.; Dukes, R.; Taylor, J.

    1975-01-01

    Results of some defect location studies on a pressure vessel are reported and correlated with those of ultrasonic inspection. Good agreement was observed, with a probability greater than 90% that a defect location detected would be confirmed by ultrasonics. This good agreement is considered to result from the use of peak sensing rather than the more commonly used leading edge triggering technique. Attention is drawn to the influence of the defect extension process upon the ease of detection, contrasting the difficulty of detecting slow crack growth with the ease of detection of pulses originating from the fracture of hard particles or their separation from the matrix, and to the influence of the Kaiser effect, which can mean that a flaw may not be detectable unless previously applied stress levels are exceeded, or that flaw growth has occurred since the previous inspection, or that some metallurgical recovery process has operated. (author)

  6. Prestressed pressure vessel for nuclear power plants

    International Nuclear Information System (INIS)

    1974-01-01

    The pressure vessel consists of a wall, a bottom, and a closure head, the wall being composed of annular segments. The closure head can be seated on the edge of the wall. Wall and closure head have got axial prestressing channels in which through-going steel tendons are arranged. They are concentrated in bundles and held above the head by anchoring devices. Within the prestressing channels of the head there are supporting jackets attached to the edge of the wall and projecting from the head through a coller. The anchoring devices, e.g. anchoring plates, may be optionally supported on the collars of the supporting jackets or on the closure head by means of auxiliary devices. The auxiliary devices for this purpose consist of extension nuts attached to the anchoring plates and closure head connecting shells. The closure head therefore may be drawn off over the anchoring devices. (DG) [de

  7. Stress intensities in flawed pressure vessels

    International Nuclear Information System (INIS)

    Smith, C.W.; Jolles, M.; Peters, W.H.

    1977-01-01

    A technique for determining the stess intensity factor (SIF) near pressure vessel flaws or cracks experimentally from photoelastic data for use in two-dimensional problems was developed in the 1950's. This technique was modified and extended to a variety of two-dimensional problems. The technique has been refined further and what has evolved may be regarded as a hybrid technique which affects a marriage between ''frozen stress'' photoelastic results and a simple least-squares digital computer program for estimating SIF values in three-dimensional problems. This technique, in its original modified form, has been shown to be applicable to a study of surface flaws and the applicability of the method to complex crack body geometries of current technological importance are discussed. The analytical foundations of the method are reviewed

  8. Design of pressure vessels. Part 2

    International Nuclear Information System (INIS)

    Grandemange, J.M.

    2008-01-01

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

  9. Holographic and acoustic emission evaluation of pressure vessels

    International Nuclear Information System (INIS)

    Boyd, D.M.

    1980-01-01

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

  10. Nuclear power plant pressure vessels. Control of piping

    International Nuclear Information System (INIS)

    2000-01-01

    The guide presents requirements for the pipework of nuclear facilities in Finland. According to the section 117 of the Finnish Nuclear Energy Degree (161/88), the Radiation and Nuclear Safety Authority of Finland (STUK) controls the pressure vessels of nuclear facilities in accordance with the Nuclear Energy Act (990/87) and, to the extent applicable in accordance with the Act of Pressure Vessels (98/73) and the rules and regulations issued by the virtue of these. In addition STUK is an inspecting authority of pressure vessels of nuclear facilities in accordance with the Pressure Vessel Degree (549/1973). According to the section of the Pressure Vessel Degree, a pressure vessel is a steam boiler, pressure container, pipework of other such appliance in which the pressure is above or may come to exceed the atmospheric pressure. Guide YVL 3.0 describes in general terms how STUK controls pressure vessels. STUK controls Safety Class 1, 2 and 3 piping as well as Class EYT (non-nuclear) and their support structures in accordance with this guide and applies the provisions of the Decision of the Ministry of Trade and Industry on piping (71/1975) issued by virtue of the Pressure Vessel Decree

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    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

  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. 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

  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. Unstable fracture of nuclear pressure vessel

    International Nuclear Information System (INIS)

    Urata, Kazuyoshi

    1978-01-01

    Unstable fracture of nuclear pressure vessel shell for light water reactors up to 1,000 MWe class is discussed in accordance with ASME Code Sec. XI. The depth of surface crack required to protect against the unstable fracture is calculated on the basis of reactor operating conditions including loss of coolant accidents. Calculated surface crack depth a is equal to tαexp(2.19(a/l)) where l is crack length and t is weld thickness. α is crack depth required to protect against the unstable fracture in terms of the ratio of crack deth to weld thickness for surface crack have infinite length. Using this α, the safety factor included for allowable defect described in Sec. XI and the effects of thickness is discussed. It is derived that allowable defect described in Sec. XI include the safety factor of two on the crack depth for crack initiation at postulated accident and the safety factor of ten for crack depth calculated from point of view of crack arrest at normal conditions. (auth.)

  17. Manufacture of an Inconel pressure vessel

    International Nuclear Information System (INIS)

    Herz, H.; Iversen, K.; Stiefelhagen, B.

    1978-01-01

    The fabrication of a thermo-shock-loaded pressure vessel of high temperature nickel alloys required the individual licensing of the basic and addition materials according to the AD data sheets Contrary to the experience of Duennbleck processars, it was found that the alloy Inconel 718 in its hardened state could not be allowed due to the formation of the brittle daves phase in the welding deposit. Positive experience was acquired however with the non-hardenable alloy Inconel 625 which could be processed as jacket materials without problem. Rods of Inconel 625 were used as similar additive for WIG welding and the same type electrode 112 for E-welding. The heat resistance required of 320 N/cm 2 at 623 0 K and the lowest notch bar value of 35 J/cm 2 at RT were well surpassed. The mixed compounds of Inconel 625 and 718 were also no problem when welding with the non-hardening additives Inconel 625 and 112 and eliminating a thermal treatment. (orig.) [de

  18. Inductive testing of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bergh, H.

    1987-01-01

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

  19. 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.

  20. 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

  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. 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.)

  3. 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

  4. Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

    International Nuclear Information System (INIS)

    Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

    1980-01-01

    An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

  5. 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

  6. High-performance fiber/epoxy composite pressure vessels

    Science.gov (United States)

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

    1978-01-01

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

  7. In-service ultrasonic inspection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Prepechal, J.; Sulc, J.

    1982-01-01

    Ultrasonic tests of pressure vessels for WWER 440 reactors, type 213 V, are carried out partly manually and partly by test equipment. The inner surface of the pressure vessel is tested using device REACTORTEST TRC which is fully mobile. The outer surface of the cylindrical parts and bottoms of the body is tested using handling equipment permanently in-built under the pressure vessel and dismountable testing heads. A set of these heads may be used for two reactor units. The testing equipment REACTORTEST TRC is equipped with a TRC 800 ultrasound device. The equipment for testing the outer surface of the vessel operates with the UDAR 16 ultrasound apparatus to which may be simultaneously connected 10 ultrasound probes and six probes for acoustic feedback. The whole system of ultrasonic tests makes possible a first-rate and reliable volume control of the whole pressure vessel and all points where cracks may originate and grow. (Z.M.)

  8. Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Bhowmik, P. K.; Shamim, J. A.; Gairola, A.; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    The study provides a brief understanding of the analysis procedure and techniques using ANSYS, such as the acceptance criteria, selection and categorization of events, thermal analysis, structural analysis including fracture mechanics assessment, crack propagation and evaluation of material properties. PTS may result from instrumentation and control malfunction, inadvertent steam dump, and postulated accidents such as smallbreak (SB) LOCA, large-break (LB) LOCA, main steam line break (MSLB), feedwater line breaks and steam generator overfill. In this study our main focus is to consider only the LB LOCA due to a cold leg break of the Optimized Power Reactor 1000 MWe (OPR1000). Consideration is given as well to the emergency core cooling system (ECCS) specific sequence with the operating parameters like pressure, temperature and time sequences. The static structural and thermal analysis to investigate the effects of PTS on RPV is the main motivation of this study. Specific surface crack effects and its propagation is also considered to measure the integrity of the RPV. This study describes the procedure for pressurized thermal shock analysis due to a loss of coolant accidental condition and emergency core cooling system operation for reactor pressure vessel.. Different accidental events that cause pressurized thermal shock to nuclear RPV that can also be analyzed in the same way. Considering the limitations of low speed computer only the static analysis is conducted. The modified LBLOCA phases and simplified geometry can is utilized to analyze the effect of PTS on RPV for general understanding not for specific specialized purpose. However, by integrating the disciplines of thermal and structural analysis, and fracture mechanics analysis a clearer understanding of the total aspect of the PTS problem has resulted. By adopting the CFD, thermal hydraulics, uncertainties and risk analysis for different type of accidental conditions, events and sequences with proper

  9. Three dimensional non-linear cracking analysis of prestressed concrete containment vessel

    International Nuclear Information System (INIS)

    Al-Obaid, Y.F.

    2001-01-01

    The paper gives full development of three-dimensional cracking matrices. These matrices are simulated in three-dimensional non-linear finite element analysis adopted for concrete containment vessels. The analysis includes a combination of conventional steel, the steel line r and prestressing tendons and the anisotropic stress-relations for concrete and concrete aggregate interlocking. The analysis is then extended and is linked to cracking analysis within the global finite element program OBAID. The analytical results compare well with those available from a model test. (author)

  10. Leakage of pressurized gases through unlined concrete containment structures

    International Nuclear Information System (INIS)

    Rizkalla, S.H.; Simmonds, S.H.

    1983-01-01

    Eight reinforced concrete specimens were fabricated and subjected to tensile membrane forces and air pressure to study the air leakage characteristics in cracked reinforced concrete members. A mathematical expression for the rate of pressurized air flowing through an idealized crack is presented. The mathematical expression is refined by using the experimental data to describe the air flow rate through any given crack pattern. Graphical charts are also presented for the calculation of the air leakage rate through concrete cracks. The concept of equivalent crack width for a given crack pattern is introduced. The mathematical expression and graphical charts are modified to include this equivalent crack width concept. The proposed technique is applicable for the prediction of the leakage from concrete containment structures or any similar structures due to high internal pressure sufficient to initiate cracking. (orig.)

  11. 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

  12. Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Bros, J.; Ballesteros, A.; Lopez, A.

    1993-01-01

    Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

  13. 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.

  14. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  15. Influence of Wind Pressure on the Carbonation of Concrete.

    Science.gov (United States)

    Zou, Dujian; Liu, Tiejun; Du, Chengcheng; Teng, Jun

    2015-07-24

    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.

  16. Leak detection device for nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    Ikeda, Jun.

    1988-01-01

    Purpose: To test the leakage of a nuclear reactor pressure vessel during stopping for a short period of time with no change to the pressure vessel itself. Constitution: The device of the present invention comprises two O-rings disposed on the flange surface that connects a pressure vessel main body and an upper cover, a leak-off pipeway derived from the gap of the O-rings at the flange surface to the outside of the pressure vessel, a pressure detection means connected to the end of the pipeway, a humidity detection means disposed to the lead-off pipeway, a humidity detection means disposed to the lead-off pipeway, and gas supply means and gas suction means disposed each by way of a check valve to a side pipe branched from the pipeway. After stopping the operation of the nuclear reactor and pressurizing the pressure vessel by filling water, gases supplied to the gap between the O-rings at the flange surface by opening the check valve. In a case where water in the pressure vessel should leak to the flange surface, when gas suction is applied by properly opening the check valve, increase in the humidity due to the steams of leaked water diffused into the gas is detected to recognize the occurrence of leakage. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  18. 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

  19. Strength-toughness requirements for thick walled high pressure vessels

    International Nuclear Information System (INIS)

    Kapp, J.A.

    1990-01-01

    The strength and toughness requirements of materials for use in high pressure vessels has been the subject of some discussion in the meetings of the Materials Task Group of the Special Working Group High Pressure Vessels. A fracture mechanics analysis has been performed to theoretically establish the required toughness for a high pressure vessel. This paper reports that the analysis performed is based on the validity requirement for plane strain fracture of fracture toughness test specimens. This is that at the fracture event, the crack length, uncracked ligament, and vessel length must each be greater than fifty times the crack tip plastic zone size for brittle fracture to occur. For high pressure piping applications, the limiting physical dimension is the uncracked ligament, as it can be assumed that the other dimensions are always greater than fifty times the crack tip plastic zone. To perform the fracture mechanics analysis several parameters must be known: these include vessel dimensions, material strength, degree of autofrettage, and design pressure. Results of the analysis show, remarkably, that the effects of radius ratio, pressure and degree of autofrettage can be ignored when establishing strength and toughness requirements for code purposes. The only parameters that enter into the calculation are yield strength, toughness and vessel thickness. The final results can easily be represented as a graph of yield strength against toughness on which several curves, one for each vessel thickness, are plotted

  20. Design of pressure vessels using shape optimization: An integrated approach

    Energy Technology Data Exchange (ETDEWEB)

    Carbonari, R.C., E-mail: ronny@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Munoz-Rojas, P.A., E-mail: pablo@joinville.udesc.br [Department of Mechanical Engineering, Universidade do Estado de Santa Catarina, Bom Retiro, Joinville, SC 89223-100 (Brazil); Andrade, E.Q., E-mail: edmundoq@petrobras.com.br [CENPES, PDP/Metodos Cientificos, Petrobras (Brazil); Paulino, G.H., E-mail: paulino@uiuc.edu [Newmark Laboratory, Department of Civil and Environmental Engineering, University of Illinois at Urbana-Champaign, 205 North Mathews Av., Urbana, IL 61801 (United States); Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 158 Mechanical Engineering Building, 1206 West Green Street, Urbana, IL 61801-2906 (United States); Nishimoto, K., E-mail: knishimo@usp.br [Department of Naval Architecture and Ocean Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil); Silva, E.C.N., E-mail: ecnsilva@usp.br [Department of Mechatronic Engineering, Escola Politecnica da Universidade de Sao Paulo, Av. Prof. Mello Moraes, 2231 Sao Paulo, SP 05508-900 (Brazil)

    2011-05-15

    Previous papers related to the optimization of pressure vessels have considered the optimization of the nozzle independently from the dished end. This approach generates problems such as thickness variation from nozzle to dished end (coupling cylindrical region) and, as a consequence, it reduces the optimality of the final result which may also be influenced by the boundary conditions. Thus, this work discusses shape optimization of axisymmetric pressure vessels considering an integrated approach in which the entire pressure vessel model is used in conjunction with a multi-objective function that aims to minimize the von-Mises mechanical stress from nozzle to head. Representative examples are examined and solutions obtained for the entire vessel considering temperature and pressure loading. It is noteworthy that different shapes from the usual ones are obtained. Even though such different shapes may not be profitable considering present manufacturing processes, they may be competitive for future manufacturing technologies, and contribute to a better understanding of the actual influence of shape in the behavior of pressure vessels. - Highlights: > Shape optimization of entire pressure vessel considering an integrated approach. > By increasing the number of spline knots, the convergence stability is improved. > The null angle condition gives lower stress values resulting in a better design. > The cylinder stresses are very sensitive to the cylinder length. > The shape optimization of the entire vessel must be considered for cylinder length.

  1. 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

  2. Recent investigations and tests with the BBR winding system for circumferential prestressing of concrete vessels and containments

    International Nuclear Information System (INIS)

    Schuett, K.; Speck, F.E.

    1993-01-01

    Prestressed concrete pressure vessels for nuclear power stations need post-tensioning systems of large capacity. For the circumferential prestressing, the continuous winding of prestressing steel has several advantages when compared to the use of large numbers of single tendons. About 15 years ago Bureau BBR Ltd (Zuerich) developed the winding system SW 8500. The further development work interrupted at that time for lack of immediate applications was resumed 4 years ago by Bureau BBR together with SUSPA on the ground of new projects being evaluated

  3. Nuclear reactor pressure vessel-specific flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.

    1992-01-01

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

  4. Crashworthy sealed pressure vessel for plutonium transport

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1980-01-01

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

  5. To the problem of reinforced concrete reactor vessel design and calculation

    International Nuclear Information System (INIS)

    Kirillov, A.P.; Artem'ev, V.P.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Paushkin, A.G.

    1980-01-01

    Modern methods for calculating reactor vessels of prestressed reinforced concrete are analyzed. It is shown that during the stage of technical and economical substantiation of reactor vessel structure for determining its stressed-deformed state engineering methods of calculation must be used, in particular, fragmentation method, method of rings and plates, and during the stages of contract and detail designs - method of finite elements and dynamic relaxation method. It is concluded that when solving cyclic symmetrical problems as well as asymmetrical problems, calculational algorithms for axis-symmetrical distributions of stresses in the vessel with provision for elastic properties of structural material may be used

  6. Nuclear Power Plant Prestressed Concrete Containment Vessel Structure Monitoring during Integrated Leakage Rate Testing Using Fiber Bragg Grating Sensors

    Directory of Open Access Journals (Sweden)

    Jinke Li

    2017-04-01

    Full Text Available As the last barrier of nuclear reactor, prestressed concrete containment vessels (PCCVs play an important role in nuclear power plants (NPPs. To test the mechanical property of PCCV during the integrated leakage rate testing (ILRT, a fiber Bragg grating (FBG sensor was used to monitor concrete strain. In addition, a finite element method (FEM model was built to simulate the progress of the ILRT. The results showed that the strain monitored by FBG had the same trend compared to the inner pressure variation. The calculation results showed a similar trend compared with the monitoring results and provided much information about the locations in which the strain sensors should be installed. Therefore, it is confirmed that FBG sensors and FEM simulation are very useful in PCCV structure monitoring.

  7. 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

  8. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    Popp, P.

    1987-01-01

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  9. Surveillance of irradiation embrittlement of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Najzer, M.

    1982-01-01

    Surveillance of irradiation embrittlement of nuclear reactor pressure vessels is briefly discussed. The experimental techniques and computer programs available for this work at the J. Stefan Institute are described. (author)

  10. 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.

  11. 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

  12. 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 ...

  13. 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

  14. 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

  15. 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

  16. Constitutive models for concrete and finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

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

    1977-01-01

    Two constitutive models for concrete are discussed. For short-term loads, the orthotropic variable modulus model is described, and for long-term loads a viscoelastic model utilizing a Dirichlet series approximation for the creep compliance function is summarized. The orthotropic variable modulus model is demonstrated in an analysis of a PCRV head with penetrations. The viscoelastic model is illustrated with a simulation of a prestressed concrete cylinder subject to non-uniform temperatures

  17. 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

  18. Assessment of the integrity of WWER type reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

  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. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    Merriman, S.H.

    1980-10-01

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

  1. Stress analysis and evaluation of a rectangular pressure vessel

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  2. Thermal annealing of an embrittled reactor pressure vessel

    International Nuclear Information System (INIS)

    Mager, T.R.; Dragunov, Y.G.; Leitz, C.

    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 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

  3. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

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

    1982-01-01

    The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

  4. Integrity of PWR pressure vessels during overcooling accidents

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  5. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  6. Structural features and in-service inspection of the LTHR-200 pressure vessel

    International Nuclear Information System (INIS)

    Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

    1993-01-01

    LTHR-200 is a low temperature district-heating reactor. It adopts double-shell design pressure vessel and metal containment. Because of the safety and structural features of the reactor, the in-service inspection of the pressure vessel can be simplified greatly. LTHR-200 is an integrated arrangement. Both its core components and the main heat exchangers are contained in the reactor pressure vessel. The coolant of the main loop is run by a full-power natural circulation and there need no main pumps and pipes. Thus, the reactor pressure vessel constitutes the pressure boundary of the reactor's main loop coolant. In regard to these features, a small-sized containment is designed for the reactor. The metal safety container with a small volume is placed closely around the reactor pressure vessel. Outside the metal containment, there is a large reinforced concrete construction for the reactor. Their main operation and design parameters are as follows: The pressure vessel: operation pressure = 2.4 MPa; design pressure = 3.0 MPa; design temperature = 250 deg C; 40 year fast neutron (E>1MeV) fluence in the belt-line region = < 10E16n/cm; internal diameter = 5000 mm; material SA516-70; shell thickness 65 mm; The metal containment: maximum operation pressure = 1.8 MPa; design pressure = 1.8 MPa; design temperature = 250 deg. C; upper internal diameter 7000 mm; lower internal diameter = 5600 mm; material = SA516-70; shell thickness, upper part = 80 mm; lower part = 50 mm. All penetrating pipes through the pressure vessel are located at the top penetration section of the shell. All the internal diameters of penetrating pipes are less than 50 mm. Inside and outside the metal containment wall respectively, isolating valves are connected to the reactor coolant pipe which passes through the containment. These two isolating valves use different driving methods. Every penetrating part of the reactor construction uses a proper form of structure according to safety requirements

  7. PWR reactor pressure vessel failure probabilities

    International Nuclear Information System (INIS)

    Dufresne, J.; Lanore, J.M.; Lucia, A.C.; Elbaz, J.; Brunnhuber, R.

    1980-05-01

    To evaluate the rupture probability of a LWR vessel a probabilistic method using the fracture mechanics under probabilistic form has been proposed previously, but it appears that more accurate evaluation is possible. In consequence a joint collaboration agreement signed in 1976 between CEA, EURATOM, JRC Ispra and FRAMATOME set up and started a research program covering three parts: a computer code development, data acquisition and processing, and a support experimental program which aims at clarifying the most important parameters used in the COVASTOL computer code

  8. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  9. 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.

  10. Pressurized-thermal-shock experiments with thick vessels

    International Nuclear Information System (INIS)

    Bryan, R.H.; Nanstad, R.K.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.

    1986-01-01

    Information is provided on the series of pressurized-thermal-shock experiments at the Oak Ridge National Laboratory, motivated by a concern for the behavior of flaws in reactor pressure vessels having welds or shells exhibiting low upper-shelf Charpy impact energies, approx. 68J or less

  11. Weld evaluation on spherical pressure vessels using holographic interferometry

    International Nuclear Information System (INIS)

    Boyd, D.M.; Wilcox, W.W.

    1980-01-01

    Waist welds on spherical experimental pressure vessels have been evaluated under pressure using holographic interferometry. A coincident viewing and illumination optical configuration coupled with a parabolic mirror was used so that the entire weld region could be examined with a single hologram. Positioning the pressure vessel at the focal point of the parabolic mirror provides a relatively undistorted 360 degree view of the waist weld. Double exposure and real time holography were used to obtain displacement information on the weld region. Results are compared with radiographic and ultrasonic inspections

  12. 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)

  13. Pressure vessel rupture within a chamber: the pressure history on the chamber wall

    International Nuclear Information System (INIS)

    Baum, M.R.

    1989-04-01

    Generally there is a large number of pressure vessels containing high pressure gas on power stations and chemical plant. In many instances, particularly on power plant, these vessels are within the main building. If a pressure vessel were to fail, the surrounding structures would be exposed to blast loads and the forces resulting from jets of fluid issuing from the breached vessel. In the case where the vessel is in a relatively closed chamber there would also be a general overpressurisation of the chamber. At the design stage it is therefore essential to demonstrate that the plant could be safely shut down in the event of a pressure vessel failure, that is, it must be shown that the chamber will not collapse thus putting the building at risk or hazarding equipment essential for a safe shut down. Such an assessment requires the loads applied to the chamber walls, roof, etc. to be known. (author)

  14. Nickel hydrogen multicell common pressure vessel battery development update

    Science.gov (United States)

    Zagrodnik, Jeffrey P.; Jones, Kenneth R.

    1992-01-01

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

  15. In-place thermal annealing of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Server, W.L.

    1985-04-01

    Radiation embrittlement of ferritic pressure vessel steels increases the ductile-brittle transition temperature and decreases the upper shelf level of toughness as measured by Charpy impact tests. A thermal anneal cycle well above the normal operating temperature of the vessel can restore most of the original Charpy V-notch energy properties. The Amry SM-1A test reactor vessel was wet annealed in 1967 at less than 343 0 C (650 0 F), and wet annealing of the Belgian BR-3 reactor vessel at 343 0 C (650 0 F) has recently taken place. An industry survey indicates that dry annealing a reactor vessel in-place at temperatures as high as 454 0 C (850 0 F) is feasible, but solvable engineering problems do exist. Economic considerations have not been totally evaluated in assessing the cost-effectiveness of in-place annealing of commercial nuclear vessels. An American Society for Testing and Materials (ASTM) task group is upgrading and revising guide ASTM E 509-74 with emphasis on the materials and surveillance aspects of annealing rather than system engineering problems. System safety issues are the province of organizations other than ASTM (e.g., the American Society of Mechanical Engineers Boiler and Pressure Vessel Code body)

  16. Structural analysis and evaluation for the design of pressure vessel

    International Nuclear Information System (INIS)

    Arai, K.; Uragami, K.; Funada, T.; Baba, K.; Kira, T.

    1977-01-01

    For the design of pressure vessel, the detailed structural analysis such as the fatigue analysis under operating conditions is required by ASME Code or Japanese regulation. Accordingly, it should be verified by the analysis that the design of the pressure vessel is in compliance with the stress limitation defined in the Code or the regulation. However, it was apparent that the analysis is very complicated and takes a lot of time to evaluate in accordance with the Code requirements. Thereupon we developed the computer program by which we can perform the stress analysis with correctness and comparatively in a short period of design work reflecting the calculation results on detailed drawings to be used for fabrication. The computer program is controlled in combination with the system of the design work and out put list of the program can be directly used for the stress analysis report which is issued to customers. In addition to the above computer program, we developed the specific three dimensional finite element computer program to make sure of the structural integrity of the vessel head and flanges which are most complex for the analysis compared with the stress distribution measured by strain gauges on the vessel head and flange. Besides the structural analysis, the fracture mechanics analysis for the purpose of preventing the pressure vessel from the brittle fracture during heat-up and cool-down operation is also important and thereby we showed herein that the pressure vessel is in safety against the brittle fracture for the specified operating conditions. As a result of the above-mentioned analysis, the pressure vessel is designed with safety from the stand-points of the structural intensity and the fracture mechanics. (auth.)

  17. 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

  18. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

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

  19. Towards a new pressure vessel standard in the European Union

    International Nuclear Information System (INIS)

    Osweiller, F.

    1995-01-01

    Since 1990 the European Commission has been preparing a new Directive which will regulate the Pressure Equipment sector in the countries of the European Union. CEN Standards devoted to pressure vessels, piping, boilers, are currently being drawn up to complete and implement this Directive. This paper focuses on the European Unfired Pressure Vessel Standard (EPVS) which is in course of development under the responsibility of CEN/TC54. The main aspects of the Standard are outlined: general structure, materials, design, fabrication, inspection and testing. The link with the European Directive is explained in connection with regulatory aspects: conformity assessment, essential safety requirements, classes of vessels, notified bodies, EC mark, status of the standard

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

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

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

  1. Design optimization of a thin walled pressure vessel

    International Nuclear Information System (INIS)

    Sadiq, S.

    2001-01-01

    Design evaluation of a pressure vessel is not only to build confidence on its integrity but also to reduce structural weight and enhance the performance of the structure. Pressure vessel, e.g., a rocket motor not only has to withstand the high operating temperatures but it must also be able to survive the internal pressures and external aerodynamic forces and bending stresses during its operation in flight. A research program was devised to study the stresses, which are generated in a thin walled pressure vessel during actual operation and its simulation with cold testing technique, i.e., by means of hydrostatic testing employing electrical resistance strain gauges on the external surface of the cylinder. The objective of the research was to uphold the performance of the vessel by reducing its thickness from 6.09 to 5.5 mm (which of course reduces the safety factor margin from 1.8 to 1.5); thereby curtailing the overall structural weight and maintaining the efficiency of the vessel itself during its live operation. The techniques employed were hydrostatic testing, data acquisition system for obtaining data on strains from the electrical resistance strain gauges and later employing V on Mises yield criterion empirical relation to computer the stresses in hoop and longitudinal directions. (author)

  2. The Assembly and Test of Pressure Vessel for Irradiation

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-02-15

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

  3. 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

  4. Welding in repair of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Pilous, V.; Kovarik, R.

    1987-01-01

    Specific welding conditions are described in repair of the pressure vessels of nuclear reactors in operation and the effect is pointed out to of neutrons on changes in steel properties. Some of the special regulations are discussed to be observed in welding jobs. The welding methods are briefly described; the half-bead method is most frequently used. It is stressed that the defect must first be identified using a nondestructive method and the stages must be defined of the welding repair of the pressure vessel. (J.B.). 4 figs., 1 tab., 16 refs

  5. 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

  6. 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).

  7. 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).

  8. Probabilistic assessment of pressure vessel and piping reliability

    International Nuclear Information System (INIS)

    Sundararajan, C.

    1986-01-01

    The paper presents a critical review of the state-of-the-art in probabilistic assessment of pressure vessel and piping reliability. First the differences in assessing the reliability directly from historical failure data and indirectly by a probabilistic analysis of the failure phenomenon are discussed and the advantages and disadvantages are pointed out. The rest of the paper deals with the latter approach of reliability assessment. Methods of probabilistic reliability assessment are described and major projects where these methods are applied for pressure vessel and piping problems are discussed. An extensive list of references is provided at the end of the paper

  9. 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)

  10. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    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. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  11. Assessment of Ultimate Load Capacity for Pre-Stressed Concrete Containment Vessel Model of PWR Design With BARC Code ULCA

    International Nuclear Information System (INIS)

    Basha, S.M.; Singh, R.K.; Patnaik, R.; Ramanujam, S.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Ultimate load capacity assessment of nuclear containments has been a thrust research area for Indian Pressurised Heavy Water Reactor (PHWR) power programme. For containment safety assessment of Indian PHWRs a finite element code ULCA was developed at BARC, Trombay. This code has been extensively benchmarked with experimental results. The present paper highlights the analysis results for Prestressed Concrete Containment Vessel (PCCV) tested at Sandia National Labs, USA in a Round Robin analysis activity co-sponsored by Nuclear Power Engineering Corporation (NUPEC), Japan and the U.S Nuclear Regulatory Commission (NRC). Three levels of failure pressure predictions namely the upper bound, the most probable and the lower bound (all with 90% confidence) were made as per the requirements of the round robin analysis activity. The most likely failure pressure is predicted to be in the range of 2.95 Pd to 3.15 Pd (Pd= design pressure of 0.39 MPa for the PCCV model) depending on the type of liners used in the construction of the PCCV model. The lower bound value of the ultimate pressure of 2.80 Pd and the upper bound of the ultimate pressure of 3.45 Pd are also predicted from the analysis. These limiting values depend on the assumptions of the analysis for simulating the concrete-tendon interaction and the strain hardening characteristics of the steel members. The experimental test has been recently concluded at Sandia Laboratory and the peak pressure reached during the test is 3.3 Pd that is enveloped by our upper bound prediction of 3.45 Pd and is close to the predicted most likely pressure of 3.15 Pd. (authors)

  12. Ultimate internal pressure capacity of concrete containment structures

    International Nuclear Information System (INIS)

    Krishnaswamy, C.N.; Namperumal, R.; Al-Dabbagh, A.

    1983-01-01

    Lesson learned from the accident at Three-Mile Island nuclear plant has necessitated the computation of the ultimate internal pressure capacity of containment structures as a licensing requirement in the U.S. In general, a containment structure is designed to be essentially elastic under design accident pressure. However, as the containment pressure builds up beyond the design value due to a more severe postulated accident, the containment response turns nonlinear as it sequentially passes through cracking of concrete, yielding of linear plate, yielding of rebar, and yielding of post-tensioning tendon (if the containment concrete is prestressed). This paper reports on the determination of the ultimate internal pressure capacity and nonlinear behavior of typical reinforced and prestressed concrete BWR containments. The probable modes of failure, the criteria for ultimate pressure capacity, and the most critical sections are described. Simple equations to hand-calculate the ultimate pressure capacity and the nonlinear behavior at membrane sections of the containment shell are presented. A nonlinear finite element analysis performed to determine the nonlinear behavior of the entire shell including nonmembrane sections is briefly discribed. The analysis model consisted of laminated axisymmetric shell finite elements with nonlinear stress-strain properties for each material. Results presented for typical BWR concrete containments include nonlinear response plots of internal pressure versus containment deflection and strains in the liner, rebar, and post-tensioning tendons at the most stressed section in the shell. Leak-tightness of the containment liner and the effect of thermal loads on the ultimate capacity are discussed. (orig.)

  13. Simulation of the behavior of pressurized underwater concrete

    Directory of Open Access Journals (Sweden)

    Ashraf Mohamed Heniegal

    2015-06-01

    Full Text Available Under-Water Concrete (UWC contains Anti-Washout Admixtures (AWA (0.0%, 0.2%, 0.3%, 0.4% and 0.5% by weight of cement with cement contents (400, 450, 500 and 550 kg/m3. All concrete mix contains silica fume and high-range water reducing (15% and 4% respectively by weight of cement. The fine to steel slag coarse aggregate was 1:1. The concrete mix was tested for slump, slump flow, compressive strength and washout resistance using two test methods based on different principles. The first method is the plunge test CRDC61 which is widely used in North America, and the second method is the pressurized air tube which has been manufactured for this research and developed to simulate the effect of water pressure on washout resistance of underwater mix. The results of compressive strength test were compared to concrete cast underwater with that cast in air. Test results indicated that the use of an AWA facilitates the production of UWC mix with the added benefit of lower washout resistance. New technique of simulating pressurized UWC is reliable for detecting UWC properties. Adding AWA (0.3–0.5% by weight of cement makes all mix acceptable according to Japanese Society of Civil Engineers.

  14. A powerful methodology for reactor vessel pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    Boucau, J.; Mager, T.

    1994-01-01

    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). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, 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. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

  15. Evaluation Of Liner Back-pressure Due To Concrete Pore Pressure At Elevated Temperatures

    International Nuclear Information System (INIS)

    James, R.J.; Rashid, Y.R.; Liu, A.S.; Gou, B.

    2006-01-01

    GE's latest evolution of the boiling water reactor, the ESBWR, has innovative passive design features that reduce the number and complexity of active systems, which in turn provide economic advantages while also increasing safety. These passive systems used for emergency cooling also mean that the primary containment system will experience elevated temperatures with longer durations than conventional plants in the event of design basis accidents. During a Loss of Coolant Accident (LOCA), the drywell in the primary containment structure for the ESBWR will be exposed to saturated steam conditions for up to 72 hours following the accident. A containment spray system may be activated that sprays the drywell area with water to condense the steam as part of the recovery operations. The liner back-pressure will build up gradually over the 72 hours as the concrete temperatures increase, and a sudden cool down could cause excessive differential pressure on the liner to develop. For this analysis, it is assumed that the containment spray is activated at the end of the 72-hour period. A back-pressure, acting between the liner and the concrete wall of the containment, can occur as a result of elevated temperatures in the concrete causing steam and saturated vapor pressures to develop from the free water remaining in the pores of the concrete. Additional pore pressure also develops under the elevated temperatures from the non-condensable gases trapped in the concrete pores during the concrete curing process. Any buildup of this pore pressure next to the liner, in excess of the drywell internal pressure, will act to push the liner away from the concrete with a potential for tearing at the liner anchorages. This paper describes the methods and analyses used to quantify this liner back-pressure so that appropriate measures are included in the design of the liner and anchorage system. A pore pressure model is developed that calculates the pressure distribution across the concrete

  16. 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.

  17. 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.)

  18. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

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

    2006-01-01

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

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

    Science.gov (United States)

    Miller, Thomas B.; Lewis, Harlan L.

    2004-01-01

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

  20. 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

  1. Multiple cell common pressure vessel nickel hydrogen battery

    Science.gov (United States)

    Zagrodnik, Jeffrey P.; Jones, Kenneth R.

    1991-01-01

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

  2. Reliability aspects of radiation damage in reactor pressure vessel mterials

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1985-01-01

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

  3. 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.

  4. 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

  5. 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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  7. 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

  8. TORT application in reactor pressure vessel neutron flux calculations

    International Nuclear Information System (INIS)

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

    1994-01-01

    The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

  9. 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.)

  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. 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

  12. Manipulator for pressure vessel open at the top

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1985-01-01

    A manipulator is provided, which has a mast, which can be fixed inside the reactor pressure vessel with a support surrounding the mast which can be moved along the mast for a carrier, which can turn around the mast and is provided with a measuring, testing, inspection or repair device. (orig./HP) [de

  13. Apparatus for carrying out ultrasonic inspection of pressure vessels

    International Nuclear Information System (INIS)

    Dent, K.H.; Challender, R.S.

    1975-01-01

    A carriage-supported manipulator for taking an ultrasonic scanner mechanism into a coolant nozzle of a nuclear reactor pressure vessel is described. The manupulator is rotatable about the axis of the nozzle and is radially expansible to urge the scanner mechanism into a scanning position within the nozzle

  14. Design, fabrication and quality assurance of pressure vessels

    International Nuclear Information System (INIS)

    Kimura, Ichiro; Miki, Masao; Yamazaki, Tsuneji; Tanaka, Yoshikazu; Sato, Misao

    1978-01-01

    The production facilities, design and manufacturing technologies, and quality assurance in the Toyo Works, Ehime Manufactory, Sumitomo Heavy Industries, Ltd., which manufactures pressure vessels, are described, and especially the actual example of non-destructive tests is shown. The Toyo Works was completed in April, 1973, to manufacture large structures such as pressure vessels, offshore structures and bridges. The total area of the site is 535,000 m 2 , that of factory buildings is 33,600 m 2 , and the outdoor assembling yard is 114,800 m 2 . The large dry dock and main installations such as 12,000 tf hydraulic press, an annealing furnace, a heat treating furnace, a quenching tank, a horizontal boring machine, 6 m vertical lathe, various welding machines, 8 MeV X-ray apparatus, sand blasting and pickling facilities, and two 160 t cranes for shipment are arranged so as to enable smooth flow of production. The standards for chemical pressure vessels in various countries are compared, and considerably high allowable stress is adopted in Europe. The design and stress analysis of pressure vessels are carried out in accordance with ASME Section 8, Div. 1 or Div. 2. As for the materials, attention must be paid to the change of properties due to heat and strain, temper brittleness, low temperature toughness and so on. The quality assurance system must be established to observe the requirements of standards. (Kako, I.)

  15. Acoustic emission signal measurements in pressure vessel testing

    International Nuclear Information System (INIS)

    Peter, A.

    1984-01-01

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

  16. Interpretation of Strain Measurements on Nuclear Pressure Vessels

    DEFF Research Database (Denmark)

    Andersen, Svend Ib Smidt; Engbæk, Preben

    1980-01-01

    with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem...

  17. Pressure vessels supported in the soil submitted to axissymetrical loads

    International Nuclear Information System (INIS)

    Gouvea, J.P. de; Bevilacqua, L.

    1982-01-01

    A pressure vessel, spherical segment or vertical cylinder, is supported in the soil and submitted to axissymetrical loads. The soil is considered as a semi-infinite elastic solid and the support as a lattice. The method of rigidity is used. (E.G.) [pt

  18. Application of improved quality control technology to pressure vessels

    International Nuclear Information System (INIS)

    Kriedt, F.

    1985-01-01

    Within the last decade, ASME Boiler and Pressure Vessel Code Section VIII-1 instituted requirements for a formal written quality control system. The results, good and bad, of this requirement are discussed. The effects are far reaching from a national economic standpoint. Quality control technology has improved. These improvements are discussed and compared to existing requirements of the CODE. Recommended improvements are suggested

  19. 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...

  20. 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...

  1. 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

  2. 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.

  3. Interpretation of strain measurements on nuclear pressure vessels

    International Nuclear Information System (INIS)

    Andersen, S.I.; Engbaek, P.

    1979-11-01

    Selected results from strain measurements on 4 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, nozzels, 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 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 with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions. (author)

  4. 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)

  5. 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.)

  6. Analysis of cracked pressure vessel nozzles by finite elements

    International Nuclear Information System (INIS)

    Reynen, J.

    1975-01-01

    In order to assess the safety of pressure vessel nozzles, the analysis should take into account cracks. The paper describes various algorithms, their computer implementations and relative merits to define in an effective way strain energy release rates along the tip front of arbitrary 3 D cracks under arbitary load including thermal strains. These techniques are basically equivalent to substructuring techniques and consequently they can be implemented to only FEM program able to deal with the data handling problems of the substructuring technique. Examples are given carried out with a substructure version of the BERSAFE system. These examples include a corner crack in a pressure vessel nozzle loaded by internal pressure and by thermal stresses. (Auth.)

  7. Helium leak testing of large pressure vessels or subassemblies

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  8. Fabrication of toroidal composite pressure vessels. Final report

    International Nuclear Information System (INIS)

    Dodge, W.G.; Escalona, A.

    1996-01-01

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

  9. 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

  10. Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

  11. AE/flaw characterization for nuclear pressure vessels

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

    1984-01-01

    This chapter discusses the use of acoustic emission (AE) detected during continuous monitoring to identify and evaluate growing flaws in pressure vessels. Off-reactor testing and on-reactor testing are considered. Relationships for identifying acoustic emission (AE) from crack growth and using the AE data to estimate flaw severity have been developed experimentally by laboratory testing. The purpose of the off-reactor vessel test is to evaluate AE monitoring/interpretation methodology on a heavy section steel vessel under simulated reactor operating conditions. The purpose of on-reactor testing is to evaluate the capability of a monitor system to function in the reactor environment, calibrate the ability to detect AE signals, and to demonstrate that a meaningful criteria can be established to prevent false alarms. An expanded data base is needed from application testing and methodology standardization

  12. Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

    International Nuclear Information System (INIS)

    Adamec, P.

    2000-12-01

    Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

  13. 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...

  14. 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)

  15. 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)

  16. Feasibility of Rectangular Concrete Pressure Vessels for Human Occupancy

    Science.gov (United States)

    1990-07-01

    incorporated: into the s 1.71 ! -1 rule-. 20 DISTRIBUTION/ AVAILABILIT ’- ,"r ABSTRACT 2’ ABSTRACT SECURITY CLASSIFICATION 7 JUNCLASSIFIED/UNLIMITED 0 SAME AS RPT...carried the end loads. Gas cooled reactors were never very popular in the US. Domestic utilities preferred boiling water reactors that operated at...Point Tower); 1975 - 9,000 psi in Chicago ( Water Tower Place); 1984 - 10,000 psi in Seattle (Century Square Bldg.); 1987 - 10,000 psi in Toronto

  17. 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.

  18. 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.

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

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

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

  20. 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.

  1. Fourier series analysis of a cylindrical pressure vessel subjected to axial end load and external pressure

    International Nuclear Information System (INIS)

    Brar, Gurinder Singh; Hari, Yogeshwar; Williams, Dennis K.

    2013-01-01

    This paper presents the comparison of a reliability technique that employs a Fourier series representation of random axisymmetric and asymmetric imperfections in a cylindrical pressure vessel subjected to an axial end load and external pressure, with evaluations prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 Rules. The ultimate goal of the reliability technique described herein is to predict the critical buckling load associated with the subject cylindrical pressure vessel. Initial geometric imperfections are shown to have a significant effect on the calculated load carrying capacity of the vessel. Fourier decomposition was employed to interpret imperfections as structural features that can be easily related to various other types of defined imperfections. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion, which are availed in the form of a double Fourier series. Fifty simulated shells generated by the Monte Carlo technique are employed in the final prediction of the critical buckling load. The representation of initial geometrical imperfections in the cylindrical pressure vessel requires the determination of respective Fourier coefficients. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries in combination with asymmetric imperfections as a function of position within the given cylindrical shell. The probability of the ultimate buckling stress exceeding a predefined threshold stress is also calculated. The method and results described herein are in stark contrast to the “knockdown factor” approach as applied to compressive stress evaluations currently utilized in industry. Further effort is needed to improve on the current design rules regarding column buckling of large diameter pressure vessels subjected to an axial end load and external pressure designed in accordance with ASME Boiler and

  2. Determination of fast neutron fluence at WWER-1000 pressure vessel

    International Nuclear Information System (INIS)

    Valenta, V. et al.

    1989-01-01

    The influence function method is an effective tool making it possible, by means of tabulated values to rapidly perform three-dimensional calculations of fast neutron fluences for various reactor core loadings and for various nuclear power plant units. The procedure for determining the spatial dependence of the fast neutron fluences in a WWER-1000 pressure vessel is described. For this, the reactor core is divided into sufficiently fine volume elements within which the neutron source can be regarded as coordinate-independent. The influence functions point to a substantial role of sources lying at the reactor core periphery. In WWER-1000 reactors, only 1 or 2 rows of peripheral assemblies are important. The influence function method makes possible a rapid and easy determination of preconditions for the assessment of the residual lifetime of the pressure vessel based on the actual reactor core loadings. (Z.M.). 7 figs., 8 refs

  3. Single pressure vessel (SPV) nickel-hydrogen battery design

    Energy Technology Data Exchange (ETDEWEB)

    Coates, D.; Grindstaff, B.; Fox, C. [Eagle-Picher Industries, Inc., Joplin, MO (United States)

    1995-07-01

    Single pressure vessel (SPV) technology combines an entire multi-cell nickel-hydrogen (NiH{sub 2}) space battery within a single pressure vessel. SPV technology has been developed to improve the performance (volume/mass) of the NiH{sub 2} system at the battery level and ultimately to reduce overall battery cost and increase system reliability. Three distinct SPV technologies are currently under development and in production. Eagle-Picher has license to the COMSAT Laboratories technology, as well as internally developed independent SPV technology. A third technology resulted from the acquisition of Johnson Controls NiH{sub 2} battery assets in June, 1994. SPV batteries are currently being produced in 25 ampere-hour (Ah), 35 Ah and 50 Ah configurations. The battery designs have an overall outside diameter of 10 inches (25.4 centimeters).

  4. 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.

  5. Pressure vessel steels: influence of chemical composition on irradiation sensitivity

    International Nuclear Information System (INIS)

    Ghoniem, M.M.; Hammad, F.H.

    1998-01-01

    Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

  6. 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.

  7. Repairing method for shroud in reactor pressure vessel

    International Nuclear Information System (INIS)

    Watanabe, Yusuke.

    1996-01-01

    The present invention provides a method of repairing a shroud disposed in a pressure vessel of a BWR type reactor. Namely, a baffle plate is disposed on the outer surface of the lower portion of the shroud supported by a shroud support of the pressure vessel. The baffle plate is connected with a lug for securing a shroud head bolt disposed on the outer surface of an upper portion of the shroud by reinforcing members. With such a constitution, when crackings are caused in the shroud, the development of the crackings can be prevented without losing the function of securing the shroud head bolt. Further, if a material having thermal expansion coefficient lower than that of austenite stainless steel is used for the material of the reinforcing member, clamping load to be applied upon attaching the auxiliary member can be reduced. As a result, operation for the attachment is facilitated. (I.S.)

  8. Apparatus for carrying out ultrasonic inspection of pressure vessels

    International Nuclear Information System (INIS)

    Dent, K.H.; Challender, R.S.

    1975-01-01

    Apparatus is described for use in carrying out ultrasonic inspection of coolant nozzles of nuclear reactor pressure vessels. It comprises a manipulator for supporting an ultrasonic scanning transducer within the coolant nozzle. The manipulator is carried by a support located within the pressure vessel and comprises a pair of legs pivotable in caliper manner to span the base of the nozzle. Means are provided for pivoting the legs together to enable free entry of the manipulator and scanning transducer into the nozzle, and for pivoting the legs apart to bring the transducer into an operating position adjacent to the wall of the nozzle. The manipulator is rotatable within the nozzle to enable scanning of its interior surface. (U.K.)

  9. 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.

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

    International Nuclear Information System (INIS)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

    1999-01-01

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

  11. Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.

    1988-05-01

    The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates.

  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. Allowable minimum upper shelf toughness for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Zahoor, A.

    1988-01-01

    The paper develops methodology and procedure for determining the allowable minimum upper shelf toughness for continued safe operation of nuclear reactor pressure vessels. Elastic-plastic fracture mechanics analysis method based on the J-integral tearing modulus (J/T) approach is used. Closed from expressions for the applied J and tearing modulus are presented for finite length, part-throughwall axial flaw with aspect ratio of 1/6. Solutions are then presented for Section III, Appendix G flaw. A simple flaw evaluation procedure that can be applied quickly by utility engineers is presented. An attractive feature of the simple procedure is that tearing modulus calculations are not required by the user, and a solution for the slope of the applied J/T line is provided. Results for the allowable minimum upper shelf toughness are presented for a range of reactor pressure vessel thickness and heatup/cooldown rates. (orig.)

  14. 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.

  15. Fabrication of pressure vessels for nuclear power plants

    International Nuclear Information System (INIS)

    Sampaio, M.S.P. de

    1982-01-01

    The status of the technology used in the fabrication of pressure vessel for nuclear power plants and the performance of the Brazilian industry in this area are presented. The followng aspects are discussed: qualification of the industries for the supplying equipment in its requirement categories; the calculation of the components; the choice of the materials; the fabrication process; and, the destructive and nondestructive tests associated to the fabrication. (E.G.) [pt

  16. The ASME Boiler and Pressure Vessel Code: overview

    International Nuclear Information System (INIS)

    Farr, J.R.

    1987-01-01

    To become familiar with the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers, it is necessary to understand the history, organization, and operation of the Boiler Code Committee as well as to become familiar with the important aspects of each Section of the Code. This chapter will review the background and contents of the Code as well as give a review of the salient contents of most sections. (author)

  17. 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)

  18. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    Kempf, Rodolfo; Fortis, Ana M.

    2007-01-01

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

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

    International Nuclear Information System (INIS)

    Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

    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.

  20. Probabilistic study of PWR reactor pressure vessel fracture

    International Nuclear Information System (INIS)

    Dufresne, J.; Lucia, A.C.; Grandemange, J.; Pellissier-Tanon, A.

    1983-01-01

    Different methods are used to evaluate the rupture probability of a nuclear pressure vessel. On of them extrapolates to nuclear pressure vessels, data of failure found in conventional pressure vessels. The disadvantage of such an approach is that the effects of systematic changes in key parameters cannot be taken into account. For example, the influence of irradiation and the use of quality assurance programs encompassing design, fabrication and materials cannot be considered. But the most important disadvantage of this method is the limited size of the representative population and consequently the high value of the upper bound failure rate corresponding to a requested confidence level. The method used in the present work involves the development of physical models based on an understanding of the failure modes and expressing the conventional concepts of fracture mechanics in a probabilistic form; the fatigue crack growth rate, calculated for conditions of cyclic loading, the initiation of unstable crack propagation, and the possibility of crack arrest. The analysis therefore requires the statistical expression of the factors and parameters which appear in the expressions of the law of crack growth and of toughness, and also those which are used in the calculation of the stress intensity factor K 1 . All input data are entered in COVASTOL code in histogram form. This code takes into account the degree of correlation between the flaw size and the Paris' law coefficients. It computes the propagation of a given defect in a given position, and the corresponding failure probability during accidental loading

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

    International Nuclear Information System (INIS)

    Andrews, R.W.

    1975-01-01

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

  2. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  3. Cracking at nozzle corners in the nuclear pressure vessel industry

    International Nuclear Information System (INIS)

    Smith, C.W.

    1986-01-01

    Cracks in nozzle corners at the pressure boundary of nuclear reactors have been frequently observed in service. These cracks tend to form with radial orientations with respect to the nozzle central axis and are believed to be initiated by thermal shock. However, their growth is believed to be primarily due to a steady plus a fluctuating internal pressure. Due to the impracticality of fracture testing of full-scale models, the Oak Ridge National Laboratory instituted the use of an intermediate test vessel (ITV) for use in fracture testing which had the same wall thickness and nozzle size as the prototype but significantly reduced overall length and diameter. In order to determine whether or not these ITVs could provide realistic data for full-scale reactor vessels, laboratory models of full-scale boiling water reactors and ITVs were constructed and tested. After briefly reviewing the laboratory testing and correlating results with service experience, results obtained will be used to draw some general conclusions regarding the stable growth of nonplanar cracks with curved crack fronts which are the most common precursors to fracture of pressure vessel components near junctures. Use of linear elastic fracture mechanics is made in determining stress-intensity distribution along the crack fronts

  4. East/west steels for reactor pressure vessels

    International Nuclear Information System (INIS)

    Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

    1997-01-01

    The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

  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. Slideline verification for multilayer pressure vessel and piping analysis

    International Nuclear Information System (INIS)

    Van Gulick, L.A.

    1983-01-01

    Nonlinear finite element method (FEM) computer codes with slideline algorithm implementations should be useful for the analysis of prestressed multilayer pressure vessels and piping. This paper presents closed form solutions useful for validating slideline implementations for this purpose. The solutions describe stresses and displacements of an internally pressurized elastic-plastic sphere initially separated from an elastic outer sphere by a uniform gap. Comparison of closed form and FEM results evaluates the usefulness of the closed form solution and the validity of the slideline implementation used

  7. Annealing of the BR3 reactor pressure vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

    1985-01-01

    The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

  8. Topic 1. Steels for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Brynda, J.; Kepka, M.; Barackova, L.; Vacek, M.; Havel, S.; Cukr, B.; Protiva, K.; Petrman, I.; Tvrdy, M.; Hyspecka, L.; Mazanec, K.; Kupca, L.; Brezina, M.

    1980-01-01

    Part 1 of the Proceedings consists of papers on the criteria for the selection and comparison of the properties of steel for pressure vessels and on the metallurgy of the said steels, the selection of suitable material for internal tubing systems, the manufacture of high-alloy steels for WWER components, the mechanical and metallurgical properties of steel 22K for WWER 440 pressure components, and of steel 10MnNi2Mo for the WWER primary coolant circuit, and the metallographic assessment of steel 0Kh18N10T. (J.P.)

  9. 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)

  10. A new model for anisotropic damage in concrete and its application to the prediction of failure of some containment vessel

    International Nuclear Information System (INIS)

    Badel, P.-B.; Godard, V.; Leblond, J.-B.

    2005-01-01

    The aim of this paper is to propose a new model for damage in concrete structures which incorporates such complex features as damage anisotropy and asymmetry between tension and compression, while being expressed in a format well suited for numerical applications and involving a limited number of material parameters which can be determined from standard experiments. A crude version of the model involving a single tonsorial internal variable representing damage in tension, and a single material parameter, is presented first. The predictions of this simple model are satisfactory in simple tension, but not so in simple compression. As a remedy, various refinements are then introduced in a second version of the model involving an additional tonsorial or scalar internal variable representing damage in compression, and five additional material parameters. An example of determination of the model parameters using experimental stress-strain curves in simple tension and compression, plus failure envelopes in biaxial tension/compression, is presented next. The model is finally applied to the numerical prediction of the failure of some containment vessel subjected to some large internal pressure, with a comparison with calculations based on a simpler isotropic variant of the model using a single scalar damage variable. The results illustrate the relevance of models incorporating both asymmetry between tension and compression and anisotropy of damage for simulations of industrial concrete structures. (authors)

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

    DEFF Research Database (Denmark)

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

    2015-01-01

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

  12. 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)

  13. 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...

  14. 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...

  15. 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...

  16. 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...

  17. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ...) 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... 46 Shipping 7 2010-10-01 2010-10-01 false Inspection of boilers, pressure vessels, piping and...

  18. 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.)

  19. An assessment of acoustic emission for nuclear pressure vessel monitoring

    International Nuclear Information System (INIS)

    Scruby, C.B.

    1983-01-01

    Recent research has greatly improved our understanding of the basic mechanisms of deformation and fracture that generate detectable acoustic emission signals in structural steels. A critical review of the application of acoustic emission (AE) to the fabrication, proof testing and in-service monitoring of nuclear pressure vessels is presented in the light of this improved understanding. The detectability of deformation and fracture processes in pressure vessel steels is discussed, and recommendations made for improving source location accuracy and the development of quantitative source assessment techniques. Published data suggest that AE can make an important contribution to fabrication monitoring, and to the detection of defects in lower toughness materials during vessel proof testing. In high toughness materials, however, the signals generated during ductile crack growth may frequently be too weak for reliable detection. The feasibility of AE for continuous monitoring has not yet been adequately demonstrated because of high background noise levels and uncertainty about AE signal strengths from the defect growth processes that occur in service. In-service leak detection by AE shows considerable promise. It is recommended that further tests are carried out with realistic defects, and under realistic conditions of loading (including thermal shock and fatigue) and of environment. (author)

  20. Pressurization of Containment Vessels from Plutonium Oxide Contents

    International Nuclear Information System (INIS)

    Hensel, S.

    2012-01-01

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

  1. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  2. Factors in the fail safe approach to pressure vessel assessment

    International Nuclear Information System (INIS)

    Connors, D.C.; Darlaston, B.J.L.; Hellen, R.A.J.

    1978-01-01

    The 'leak before break' concept is described in the context of pressure vessel assessment. The factors which determine whether a pipe containing an axial flaw will leak or break under pressure loading are discussed using a post-yield fracture mechanics method. A model is used in which it is assumed that initially the ligament beneath the flaw fails to form a full thickness defect in the pipe. The stability of the full thickness defect at the pressure causing ligament failure is then examined to ascertain whether it would remain at the snap through length or would propagate by fast fracture, to form a leak or a break. The method is used to analyse the results of a series of pipe rupture tests, and it is found that a distinction between leaks and breaks is achieved. (author)

  3. Acoustic emission measurements at the pressure vessel ZB2

    International Nuclear Information System (INIS)

    Tirbonod, B.; Hanacek, L.

    1990-01-01

    The work presented here is the Swiss contribution to the project 'Zwischenbehaelter 2 (ZB2)' hosted by the 'Bundesministerium fuer Forschung und Technologie' of the Federal Republic of Germany. One of the crack-like defects introduced at the inside surface of the thick-walled pressure vessel ZB2 was locally monitored by acoustic emission. The measurement system was broadband (0.5 - 5 MHz) and allowed a threedimensional location of the source. The vessel was subjected to different tests. Signals were recorded during the second series of hydrotests, fast pressure cycles and fatigue test at 50 C. About 1 signal per hydrotest or cycle was recorded. For the hydrotests the signals were recorded generally at loading in the intermediate range of pressure; the sources were located in the artificial defect. Recurrent and non recurrent signals were recorded during the fatigue test. At loading, signals were captured up to the maximum pressure and for the recurrent signals at well defined pressure ranges. All the sources (except one, located in the base material ahead of the artificial defect) were situated in the artificial defect. The pressure and location depended on the loading phase and on the cycle range. The measurements were discussed by describing the signals by measurement, signal and source parameters. The goal was to identify the source mechanism and to assess the growth of the defect. For the hydrotests the identification of the mechanism at loading remains open. For the fatigue test the source situated in the base material was attributed to a primary mechanism; this source could assess the growth of the defect on the basis of linear elastic fracture mechanics. A secondary mechanism was suggested for recurrent sources active at loading. For all the tests, the sources active at unloading were attributed to a secondary mechanism. (author)

  4. Experimental and theoretical studies on the high pressure vessel

    International Nuclear Information System (INIS)

    So, Dong Sup

    1992-02-01

    A High Pressure Melt Ejection (HPME) is one of the most important phenomena relevant to Direct Containment Heating(DCH) which could lead to an early containment failure in a several accident of PWRs. Dispersal of core debris following a postulated high pressure failure of PWR reactor vessel has been investigated by experimental works and one-dimensional computer modeling to find the relation between the fraction of melt simulant retained in the cavity and the reactor vessel initial conditions as well as to examine the hydrodynamic processes in a reactor cavity geometry. Simulated HPME experiments have been performed with two small-scale (1/25-th and 1/41-st) transparent reactor cavity models of the Young-Gwang unit 1 and 2. Wood's metal and water have been used as melt sumulants while high pressure nitrogen and carbon dioxide have been used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. The high speed movies of the transient tests showed that no fraction of the melt simulant exits the cavity model via the vertical cavity tunnel under its own momentum, and that the discharged simulant from the pressure vessel exits the reactor cavity model during the gas blowdown. The principal removal mechanism seemed to be a combined mechanism of film entrainment and particle levitation due to the driving force of the blowdown gas. Experimental data for the fraction of melt simulant retained in the cavity model (Y f ) during a postulated scenario of the HPME from PWR pressure vessels have been obtained as a function of various test parameters. These data have been used to develop a correlation for Y f that fits all the data (a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Y f are determined based on the correlation obtained here and

  5. Tribology aspects of a pressure vessel closure subjected to pressure cycling

    International Nuclear Information System (INIS)

    George, A.F.; Williams, M.E.

    1988-04-01

    A repair method being considered for a steel pressure vessel is to cut away the faulty part leaving an unreinforced circular hole in the curved wall and cover it with a sealed plate placed inside. In order to investigate the structural properties of such a repair a large model vessel (6m by 2m) was tested under pressure (about 2.5 MPa) and pressure cycling. This cycling caused relative movements at the loaded interface between the lid and the vessel. A tribological examination of the rubbing surfaces was carried out. The tribological examination is described and a small supporting programme of laboratory scaling tests. It gives the results and attempts to interpret them with particular attention given to wear, fretting fatigue and scaling to plant conditions. (author)

  6. 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

  7. On the state of acoustic emission analysis in pressure vessel and model vessel testing

    International Nuclear Information System (INIS)

    Morgner, W.; Theis, K.; Henke, F.; Imhof, D.

    1985-01-01

    In the GDR acoustic emission analysis is being applied primarily in connection with hydraulic pressure testing of vessels in chemical industry. It is, however, also used for testing and monitoring of equipment and components in other branches of industry. The state-of-the-art is presented with regard to equipment needed, training of personnel, licensing of testing methods and appropriate testing procedures. In particular, the evaluation of the sum curves and amplitude distributions is explained, using rupture tests of two oxygen cylinders and a compressed-air bottle as examples. (author)

  8. 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

  9. The Clementine Nickel Hydrogen Common Pressure Vessel Battery

    OpenAIRE

    Garner, Christopher

    1994-01-01

    The Clementine spacecraft was launched in January 1994 to demonstrate advanced lightweight technologies for the Ballistic Missile Defense Organization (BMDO). One of the key technologies was the first use of a multi-cell nickel hydrogen (NiH2) common pressure vessel (CPV) battery. The 5.0 inch diameter, 22 cell, 15.0 ampere-hour NiH2 CPV battery was manufactured by Johnson Controls Battery Group Inc., (JCBGI). Battery test and integration was performed by the Naval Research Laboratory (NRL). ...

  10. 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.

  11. 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)

  12. Design criteria and pressure vessel codes - an American view

    International Nuclear Information System (INIS)

    Tuppeny, W.H.

    1975-01-01

    To the pressure vessel designer, codes and criteria represent the common ground where the stress analyst and the metallurgist must interact and evolve rules and procedures which will ensure safety and open-ended responsiveness to technological, economic, and environmental change. The paper briefly discusses the evolution and rationale behind the current ASME code sections -emphasizing those portions applicable to designs operating in the creep range. The author then proposes a plan of action so that the analysts and materials people can make optimum use of time and resources, and evolve data and design criteria which will be responsive to changing technology and the economic and safety requirements of the future. (author)

  13. Evaluation of flaws or service induced cracks in pressure vessels

    International Nuclear Information System (INIS)

    Riccardella, P.C.; Copeland, J.F.; Gilman, J.

    1987-01-01

    An overview of the ASME flaw evaluation procedures for nuclear pressure vessels is presented, with emphasis on fatigue crack growth evaluations. Environmental and load-rate effects are further considered with respect to new crack growth data and a time-dependent crack growth model. This new crack growth model is applied to evaluate feedwater nozzle cracking in boiling water reactors and is compared to current and past ASME crack growth curves. The time-dependent model bounds the observed cracking and indicates that more detailed consideration of material susceptibility, in terms of sulfur content and product form, is needed

  14. Neutron irradiation effects in pressure vessel steels and weldments

    Energy Technology Data Exchange (ETDEWEB)

    Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

    1994-12-31

    This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

  15. Automated ultrasonic shop inspection of reactor pressure vessel forgings

    International Nuclear Information System (INIS)

    Farley, J.M.; Dikstra, B.J.; Hanstock, D.J.; Pople, C.H.

    1986-01-01

    Automated ultrasonic shop inspection utilizing a computer-controlled system is being applied to each of the forgings for the reactor pressure vessel of the proposed Sizewell B PWR power station. Procedures which utilize a combination of high sensitivity shear wave pulse echo, 0 degrees and 70 degrees angled longitudinal waves, tandem and through-thickness arrays have been developed to provide comprehensive coverage and an overall reliability of inspection comparable to the best achieved in UKAEA defect detection trials and in PISC II. This paper describes the ultrasonic techniques, the automated system (its design, commissioning and testing), validation and the progress of the inspections

  16. Apparative developments for inservice inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Bohn, H.; Ruthrof, K.; Barbian, O.A.; Kappes, W.; Neumann, R.; Stanger, H.K.

    1987-01-01

    Emphasizing PWR pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implenentation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system. (orig./HP)

  17. Possibility logic applied to pressure vessel residual lifetime prediction

    International Nuclear Information System (INIS)

    Garribba, S.; Lucia, A.C.; Volta, G.

    1985-01-01

    The adequacy is discussed of a probability measure to deal with the different types of uncertainty affecting any pressure vessel lifetime prediction. A more comprehensive framework derived from the fuzzy set theory and including as particular case possibility and probability measures is considered. With reference to the most critical step of lifetime assessment (the ND inspection), the paper compares the results, obtained adopting a possibility measure or a probability measure, in the representation models, fault tree-event tree, and in the decision models

  18. Electrode for welding steel for WWER-1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    Of two types of electrodes, ie., with an alloyed core and with an unalloyed core, an electrode was chosen consisting of a basic coat and an unalloyed core. Fluctuations are shown of shear strength, tensile strenght and contraction with the welding mode and annealing temperature. It was found that pre-heating to 250 and 350 degC, respectively, was most suitable for welding a pressure vessel manufactured from material designated SKODA A3/II. Annealing aimed at removing stress was chosen at 650 to 700 degC. (H.S.)

  19. 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.

  20. 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)

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

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

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

  2. 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

  3. 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

  4. 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%)

  5. 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.

  6. 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.

  7. 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

  8. Use of superheated steam to anneal the reactor pressure vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1994-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 degrees F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 degrees F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors

  9. Pressure vessel codes: Their application to nuclear reactor systems

    International Nuclear Information System (INIS)

    1966-01-01

    A survey has been made by the International Atomic Energy Agency of how the problems of applying national pressure vessel codes to nuclear reactor systems have been treated in those Member States that had pressurized reactors in operation or under construction at the beginning of 1963. Fifteen answers received to an official inquiry form the basis of this report, which also takes into account some recently published material. Although the answers to the inquiry in some cases data back to 1963 and also reflect the difficulty of describing local situations in answer to standard questions, it is hoped that the report will be of interest to reactor engineers. 21 refs, 1 fig., 2 tabs

  10. Probabilistic structural integrity of reactor vessel under pressurized thermal shock

    International Nuclear Information System (INIS)

    Myung Jo Hhung; Young Hwan Choi; Hho Jung Kim; Changheui Jang

    2005-01-01

    Performed here is a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel. A round robin consisting of 1 prerequisite study and 5 cases for probabilistic approaches is proposed, and all organizations interested are invited. The problems are solved and their results are compared to issue some recommendation of best practices in this area and to assure an understanding of the key parameters of this type of approach, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria. Six participants from 3 organizations in Korea responded to the problem and their results are compiled in this study. (authors)

  11. 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.

  12. 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)

  13. Analytical and experimental vibration analysis of BWR pressure vessel internals

    International Nuclear Information System (INIS)

    Krutzik, N.; Schad, O.

    1975-01-01

    This report attempts to evaluate the validity as well as quality of several analytical methods in the light of presently available experimental data for the internals of pressure vessels of boiling-water-reactor-types. The experimental checks were performed after the numerical analysis was completed and showed the accuracy of the numerical results. The analytical investigations were done by finite element programmes - 2-dimensional as well as 3-dimensional, where the effect of the mass distribution with parts of virtual masses on the dynamic response could be studied in depth. The experimental data were collected at various different plants and with different mass correlations. Besides evaluating the dynamic characteristics of the components, tests were also performed to evaluate the vibrations of the pressure vessel relative to the main structure. After analysing extensive recorded data much better understanding of the response under a variety of loading- and boundary conditions could be gained. The comparison of the results of analytical studies with the experimental results made a broad qualitative evaluation possible. (Auth.)

  14. Assessment of environmentally assisted cracking in PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Tice, D.R.

    1991-01-01

    There is a possibility that extension of pre-existing flaws in the reactor pressure vessel of a pressurised water reactor (PWR) may occur by environmentally assisted cracking, in particular by corrosion fatigue under cyclic transient loading. Crack growth predictions have usually been carried out using cyclic crack growth rate (da/dN) versus stress intensity range (δK) curves, such as those given in Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code. However, the inherent time dependent nature of environmental cracking processes renders such an approach unrealistic. The present paper describes the development of an alternative time based assessment methodology. Illustrative calculations of expected crack growth of assumed defects made using the cyclic (ASME XIA) and time-based approaches are compared. The results illustrate that crack growth predicted by the time-based approach can be greater or less than that calculated by the traditional method. For a PWR operated with good control of water chemistry, actual crack growth rates are expected to be well below those predicted by the ASME code. (Author)

  15. Dynamic fracture characterization of a pressure vessel steel

    International Nuclear Information System (INIS)

    Schmitt, W.; Boehme, W.; Klemm, W.; Memhard, D.; Winkler, S.

    1991-01-01

    Dynamic events are characterized by time and space-dependent stress and strain fields caused by wave or inertia effect. The dynamic effect at cracks may be originated from the rapid loading rate or impact loading of a structure containing a stationary crack or the time-dependent stress and strain fields of a propagating or arresting crack itself. Dynamic effects complicate the analysis of crack tip stress and strain fields, and usually considerable experimental effort and numerical technique are required. High loading rate influences the deformation and yield behavior and also the fracture toughness of materials. In order to know the propagation and arrest behavior of cracks, a heat of a German reactor pressure vessel steel was investigated, and the dynamic J-resistance curves were evaluated with large three-point bending specimens by impact loading, moreover, the crack propagation energy at large crack extension was determined with wide tension plates. The material tested was a ferritic pressure vessel steel, ASTM A 508 Cl 2. The dynamic J-resistance curves and numerical simulation and fractographic examination, and crack propagation energy are reported. (K.I.)

  16. Remote controlled stud bolt handling device for reactor pressure vessel

    International Nuclear Information System (INIS)

    Shindo, Takenori; Shigehiro, Katsuya; Ito, Morio; Okada, Kenji

    1988-01-01

    In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

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

    International Nuclear Information System (INIS)

    Brumovsky, M.; Polachova, H.

    1995-01-01

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

  18. Compressed natural gas transportation by utilizing FRP composite pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

    This paper discussed the Trans Ocean Gas (TOG) method for transporting compressed natural gas (CNG). As demand for natural gas increases and with half of the world's reserves considered stranded, a method to transport natural gas by ship is needed. CNG transportation is widely viewed as a viable method. Transported as CNG, stranded gas reserves can be delivered to existing markets or can create new natural gas markets not applicable to liquefied natural gas (LNG). In contrast to LNG, compressed gas requires no processing to offload. TOG proposes that CNG be transported using fiber reinforced plastic (FRP) pressure vessels which overcome all the deficiencies of proposed steel-based systems. FRP pressure vessels have been proven safe and reliable through critical applications in the national defense, aerospace, and natural gas vehicle industries. They are light-weight, highly reliable, have very safe failure modes, are corrosion resistant, and have excellent low temperature characteristics. Under TOG's scheme, natural gas can be stored at two thirds the density of LNG without costly processing. TOG's proposed design and testing of a CNG system was reviewed in detail. 1 fig.

  19. Reassessment of PWR pressure-vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Ball, D.G.

    1983-01-01

    A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

  20. 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)

  1. 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)

  2. Structural integrity test of prestressed concrete containment vessel for Tsuruga Unit No. 2 Nuclear Power Station

    International Nuclear Information System (INIS)

    Tamura, S.; Nagata, K.; Takeda, T.; Yamaguchi, T.; Nakayama, T.

    1987-01-01

    In introducing the PCCV to Japan, various verification tests were carried out to understand the structural performance of the PCCV and confirm the reliability of its design. In addition to those tests, a Structural Integrity Test (SIT) was conducted in Feb. 1986 as a final acceptance test. This report discusses the results of the SIT on the PCCV. The test was carried out simultaneously with an Integrated Leak Rate Test (ILRT) under the same pressure sequence. 1) Pressure-displacement relationships and pressure-strain relationships were more or less linear. 2) The measured displacement values at the maximum pressure (4.5 kgf/cm 2 G) corresponded well with calculated values. Correspondence with converted displacement obtained from strain and measured displacement was also good. 3) The residual displacement when 24 hours had elapsed after completion of depressurization was not more than 10% of the displacement at the maximum pressure. 4) The variation in tendon force at the maximum pressure is smaller than the calculated value in proportion to the elongation of the PCCV. 5) Although fine surface cracks due to shrinkage of concrete were seen, new structural cracks due to pressure were not observed. The leakage rate was evaluated at 0.016% of volume per day. It is much smaller than the design value of 0.1% of volume per day. (orig./HP)

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

    Science.gov (United States)

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

    2006-01-01

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

  4. Prevention against fragile fracture in PWR pressure vessel in the presence of pressurized thermal shock

    International Nuclear Information System (INIS)

    Carmo, E.G.D. do; Oliveira, L.F.S. de; Roberty, N.C.

    1984-01-01

    A method for the determination of operational limit curves (primary pressure versus temperature) for PWR is presented. Such curves give the operators indications related to the safety status of the plant concerning the possibility of a pressurized thermal shock. The method begins by a thermal analysis for several postulated transients, followed by the determination of the thermomechanical stresses in the vessel and finally it makes use of the linear elasticity fracture mechanics. Curves are shown for a typical PWR. (Author) [pt

  5. Material properties characterization - concrete

    International Nuclear Information System (INIS)

    England, G.L.; MacLeod, J.S.

    1978-01-01

    A review is presented of the six contributions in the SMiRT 4 conference to Session H5 on structural analysis of prestressed concrete reactor pressure vessels. These relate to short term stress-strain aspects of concrete loaded beyond the linear range in uniaxial and biaxial stress fields, to some time and temperature dependent properties of concrete at working stress levels, and to a programme of strain-gauge testing for the assessment of concrete properties. From the information discussed, it is clear that there are difficulties in determining material properties for concrete, and these are summarised. (UK)

  6. Reactor pressure vessel failure probability following through-wall cracks due to pressurized thermal shock events

    International Nuclear Information System (INIS)

    Simonen, F.A.; Garnich, M.R.; Simonen, E.P.; Bian, S.H.; Nomura, K.K.; Anderson, W.E.; Pedersen, L.T.

    1986-04-01

    A fracture mechanics model was developed at the Pacific Northwest Laboratory (PNL) to predict the behavior of a reactor pressure vessel following a through-wall crack that occurs during a pressurized thermal shock (PTS) event. This study, which contributed to a US Nuclear Regulatory Commission (NRC) program to study PTS risk, was coordinated with the Integrated Pressurized Thermal Shock (IPTS) Program at Oak Ridge National Laboratory (ORNL). The PNL fracture mechanics model uses the critical transients and probabilities of through-wall cracks from the IPTS Program. The PNL model predicts the arrest, reinitiation, and direction of crack growth for a postulated through-wall crack and thereby predicts the mode of vessel failure. A Monte-Carlo type of computer code was written to predict the probabilities of the alternative failure modes. This code treats the fracture mechanics properties of the various welds and plates of a vessel as random variables. Plant-specific calculations were performed for the Oconee-1, Calvert Cliffs-1, and H.B. Robinson-2 reactor pressure vessels for the conditions of postulated transients. The model predicted that 50% or more of the through-wall axial cracks will turn to follow a circumferential weld. The predicted failure mode is a complete circumferential fracture of the vessel, which results in a potential vertically directed missile consisting of the upper head assembly. Missile arrest calculations for the three nuclear plants predict that such vertical missiles, as well as all potential horizontally directed fragmentation type missiles, will be confined to the vessel enclosre cavity. The PNL failure mode model is recommended for use in future evaluations of other plants, to determine the failure modes that are most probable for postulated PTS events

  7. 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

  8. Fracture behaviour assessment of a flawed pressure vessel in the hydro-test

    Energy Technology Data Exchange (ETDEWEB)

    Sarkimo, M; Rintamac, R

    1988-12-31

    This document deals with the fracture properties of a flawed pressure vessel. The experiment was carried out within the Nordic Countries on a vessel in a Finnish refinery. The instrumentation used included acoustic emission. Some results are provided. (TEC).

  9. A-508 class 3 forgings for pressure vessels

    International Nuclear Information System (INIS)

    Comon, J.

    1977-01-01

    The manufacture of the forged parts of the first PWR nuclear pressure vessel installed in France started in the Creusot-Loire's Forge Plant in 1961. Since this date, more than 300 forgings of this type were delivered (flanges, rings, zones, and nozzles). The major part of these forgings were made of Mn, Ni, Mo steel (SA 508 class 3). They represent a population large and homogeneous enough to attempt a statistical analysis of chemical and mechanical test results. The aim of this analysis was double: (1) a better knowledge of the scattering of the results and a better estimate of what can be introduced or accepted in a specification, and (2) the setting up of correlations existing between these results, particularly between chemical analysis and mechanical test results. In addition to this statistical analysis concerning industrial results, several laboratory studies are presented, giving a more complete characterization of SA 508 class 3. All these results form a very complete documentation showing that SA 508 class 3 steel is suitable for the manufacture of large forged vessels requiring a high degree of reliability

  10. Consequence evaluation of hypothetical reactor pressure vessel support failure

    International Nuclear Information System (INIS)

    Lu, S.C.; Holman, G.S.; Lambert, H.E.

    1991-01-01

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

  11. Recent development for inservice inspection of reactor pressure vessels

    International Nuclear Information System (INIS)

    Fischer, K.; Engl, G.; Rathgeb, W.; Heumueller, R.

    1991-01-01

    The German Nuclear Code (KTA-rules) requires a full scope inservice inspection (ISI) of reactor pressure vessels within a period of four years. This has a remarkable influence on plant operation and economy. Therefore, the development of advanced inspection equipment and techniques is directed not only to the enhancement of defect detectability and flaw sizing capabilities but also to reducing inspection times. A new manipulator system for PWR vessels together with fast data processing reduces the time for ISI of modern RPVs to 7 days. A new multichannel UT-system based on ALOK principle offers increased ultrasonic information with comfortable and rapid evaluation and presentation of results together with enhanced sizing capabilities. For specific inspection problems characterized by geometrical complexity the application of phased array probes in connection with UT-tomography provides improved ultrasonic information together with a streamlined manipulator principle and simplification of set up and tear down at the component which results in considerable reduction of radiation exposure. (orig.)

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

  13. 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.

  14. 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

  15. Italian developments in the ultrasonic examination of pressure vessels

    International Nuclear Information System (INIS)

    Regis, V.

    1987-01-01

    A review of developments being pursued in Italy in ultrasonics for application to pressure vessels is presented. Although nuclear construction in Italy has suffered heavy delays, R and D activities promoted by the Italian Electricity Board in the mid 1970s on advanced UT for non-destructive inspection of thick welded sections made it possible to obtain significant results scored by CISE Laboratories, mainly through the design, construction and qualification of the manual UT spectroscopy and signal processing computerized ARICE system and of the mechanized multifrequency acoustic holography HADIS system. Meanwhile theoretical ultrasonic modelling is actively studied in order to implement software applications and the overall reliability of UT inspections with regard to flaw detection, location and sizing. Selected contributions from manufacturers and service companies with a view to improving UT practice are acknowledged, and still wider technology transfers may be expected in the future, also under ENEA industrial promotion programmes. (author)

  16. Effect of aging on properties of pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Druce, S.G.; Gage, G.; Jordan, G.

    1986-04-01

    Manganese-molybdenum-nickel steels are used in nuclear pressure vessels operating at temperatures up to 350/sup 0/C. The effects of thermal ageing in the temperature range 300-550/sup 0/C for durations up to 2 x 10/sup 4/ h have been studied in conventionally quenched and tempered and simulated heat-affected-zone (HAZ) microstructural conditions. Quantitative fractography and Auger spectroscopy have been used to relate changes in mechanical properties with changes in fracture mode and grain boundary chemistry. Aging increases the ductile-brittle transition temperature by an amount dependent on material, prior heat treatment, aging temperature and time. Embrittlement is associated with segregation of phosphorus to grain boundaries and is modelled using McLean's approach to equilibrium segregation.

  17. 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.

  18. 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

  19. 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.

  20. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    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.

  2. 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)

  3. 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

  4. Preparation of the Shippingport reactor pressure vessel shipping package

    International Nuclear Information System (INIS)

    Yannitell, D.M.

    1988-01-01

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

  5. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  6. 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

  7. 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

  8. The behavior of shallow flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Rolfe, S.T.

    1991-11-01

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

  9. 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...

  10. 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...

  11. 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

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

    Science.gov (United States)

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

    2013-12-01

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

  13. 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.)

  14. 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)

  15. 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,...

  16. ZZ SAIL, Albedo Scattering Data Library for 3-D Monte-Carlo Radiation Transport in LWR Pressure Vessel

    International Nuclear Information System (INIS)

    1982-01-01

    1 - Description of problem or function: Format: SAIL format; Number of groups: 23 neutron / 17 gamma-ray; Nuclides: Type 04 Concrete and Low Carbon Steel (A533B). Origin: Science Applications, Inc (SAI); Weighting spectrum: yes. SAIL is a library of albedo scattering data to be used in three-dimensional Monte Carlo codes to solve radiation transport problems specific to the reactor pressure vessel cavity region of a LWR. The library contains data for Type 04 Concrete and Low Carbon Steel (A533B). 2 - Method of solution: The calculation of the albedo data was perform- ed with a version of the discrete ordinates transport code DOT which treats the transport of neutrons, secondary gamma-rays and gamma- rays in one dimension, while maintaining the complete two-dimension- al treatment of the angular dependence

  17. FAVOR: A new fracture mechanics code for reactor pressure vessels subjected to pressurized thermal shock

    International Nuclear Information System (INIS)

    Dickson, T.L.

    1993-01-01

    Probabilistic fracture mechanics (PFM) analysis is a major element of the comprehensive probabilistic methodology endorsed by the Nuclear Regulatory Commission (NRC) for evaluation of the integrity of pressurized water reactor pressure vessels subjected to pressurized-thermal-shock (PTS) transients. OCA-P and VISA-II are PTS PFM computer codes that are currently referenced in Regulatory Guide 1.154 as acceptable codes for performing plant-specific analyses. These codes perform PFM analyses to estimate the increase in vessel failure probability as the vessel accumulates radiation damage over the operating life of the vessel. Experience with the application of these codes in the last few years has provided insights into areas where they could be improved. As more plants approach the PTS screening criteria and are required to perform plant-specific analyses, there will be an increasing need for an improved and validated PTS PFM code that is accepted by the NRC and utilities. The NRC funded Heavy Section Steel Technology Program (HSST) at the Oak Ridge National Laboratory is currently developing the FAVOR (Fracture Analysis of Vessels: Oak Ridge) code, which is expected to meet this need. The FAVOR code incorporates the most important features of both OCA-P and VISA-II and contains some new capabilities such as (1) a PFM global modeling methodology; (2) the calculation of the axial stress component associated with coolant streaming beneath an inlet nozzle; (3) a library of stress intensity factor influence coefficients, generated by the NQA-1 certified ABAQUS computer code, for an appropriate range of two and three dimensional inner-surface flaws; (4) the flexibility to generate a variety of output reports; and (5) enhanced user friendliness

  18. Analysis of aging mechanism and management for HTR-PM reactor pressure vessel

    International Nuclear Information System (INIS)

    Sun Yunxue; Shao Jin

    2015-01-01

    Reactor pressure vessel is an important part of the reactor pressure boundary, its important degree ranks high in ageing management and life assessment of nuclear power plant. Carrying out systematic aging management to ensure reactor pressure vessel keeping enough safety margins and executing design functions is one of the key factors to guarantee security and stability operation for nuclear power plant during the whole lifetime and prolong life. This paper briefly introduces the structure and aging mechanism of reactor pressure vessel in pressurized water reactor nuclear power plant, and introduces the design principle and structure characteristics of HTR-PM. At the same time, this paper carries out preliminary analysis and exploration. and discusses aging management of HTR-PM reactor pressure vessel. Finally, the advice of carring out aging management for HTR-PM reactor pressure vessel is proposed. (authors)

  19. Analysis of cracked pressure vessel nozzles by finite elements

    International Nuclear Information System (INIS)

    Reynen, J.

    1975-01-01

    The paper describes various algorithms, their computer implementations and relative merits to define in an effective way strain energy release rates along the tip front of arbitrary 3D cracks under arbitrary load including thermal strains. These techniques are basically equivalent to substructuring techniques and consequently they can be implemented to any FEM program able to deal with the data handling problems of the substructuring technique. Special finite elements with a built-in stress-singularity are not necessary although their use contributes to accuracy and the mesh can be coarser. Examples are given carried out with a substructure version of the BERSAFE system. These examples include a corner crack in a pressure vessel nozzle loaded by internal pressure and by thermal stresses. Although not of any fundamental importance, in practice the difficulties consist in generating an appropriate mesh to represent the crack front. For the example of the corner crack in a nozzle the problem has been solved by developing a special purpose mesh generation program (EURCRACK)

  20. Dynamic loads on reactor vessel components by low pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-01-01

    Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de